ge-hitachi abwr design control document tier 1 & 2, rev. 4

ABWRDesignControlDocument

GE Nuclear Energy

Rev. 4March 1997

@

i

Rev. 3

Design Control Document

ABWR

Effective Pages of the Design Control Document ................................................................. Volume 1

Introduction to the Design Control Document .................................................................... Volume 1

Tier 1, Section 1.0 Introduction ............................................................................................ Volume 1

Tier 1, Section 2.0 Certified Design Material for ABWR Systems......................................... Volume 1

Tier 1, Section 3.0 Additional Certified Design Material..................................................... Volume 2

Tier 1, Section 4.0 Interface Requirements ........................................................................... Volume 2

Tier 1, Section 5.0 Site Parameters......................................................................................... Volume 2

Tier 2, Chapter 1 Introduction and General Plant Description of Plant............................ Volume 3

Tier 2, Chapter 2 Site Characteristics.................................................................................... Volume 3

Tier 2, Chapter 3 Design of Structures, Conponents, Equipment and Systems..........Volumes 4,5,6

Tier 2, Chapter 4 Reactor........................................................................................................ Volume 7

Tier 2, Chapter 5 Reactor Coolant System and Connected Systems .................................... Volume 7

Tier 2, Chapter 6 Engineered Safety Features ....................................................................... Volume 8

Tier 2, Chapter 7 Instrumentation and Control Systems...................................................... Volume 9

Tier 2, Chapter 8 Electric Power ............................................................................................ Volume 9

Tier 2, Chapter 9 Auxiliary Systems.........................................................................Volumes 10, 11, 12

Tier 2, Chapter 10 Steam and Power Conversion System................................................... Volume 13

Tier 2, Chapter 11 Radioactive Waste Management ........................................................... Volume 13

Tier 2, Chapter 12 Radiation Protection.............................................................................. Volume 13

Tier 2, Chapter 13 Conduct of Operations.......................................................................... Volume 14

Tier 2, Chapter 14 Intial Test Program................................................................................ Volume 14

Tier 2, Chapter 15 Accident and Analysis ............................................................................ Volume 15

Tier 2, Chapter 16 Technical Specifications.....................................................Volumes 16, 17, 18, 19

Tier 2, Chapter 17 Quality Assurance .................................................................................. Volume 20

Tier 2, Chapter 18 Human Factors Engineering................................................................. Volume 20

Tier 2, Chapter 19 Response to Severe Accident Policy Statement.......................Volumes 21, 22, 23

Tier 2, Chapter 20 Question and Resonse Guide.........................................................Volumes 24, 25

Tier 2, Chapter 21 Engineering Drawings ......................................................Volumes 26 through 31

Design Control Document Table Of Contents

/ii

Table of Contents i

Rev. 1

Design Control Document/Tier 2ABWR

1.0 Introduction and General Description of Plant1.1 Introduction1.2 General Plant Description1.3 Comparison Tables1.4 Identification of Agents and Contractors1.5 Requirements for Further Technical Information1.6 GE Topical Reports and Other Documents1.7 Drawings1.8 Conformance with Standard Review Plan and Applicability of Codes and Standards1.9 COL License Information1A Response to TMI Related Matters1AA Plant Shielding to Provide Access to Vital Areas1B Not Used1C ABWR Station Blackout Considerations

2.0 Site Characteristics2.1 Limits Imposed on SRP Section II Acceptance Criteria by ABWR Standard Plant2.2 Requirements for Determination of ABWR Site Acceptability2.3 COL License Information2A Input to CRAC 2 Computer Code for Determination of ABWR Site Acceptability

3.0 Design of Structures, Components, Equipment and Systems3.1 Conformance with NRC General Design Criteria3.2 Classification of Structures, Components, and Systems3.3 Wind and Tornado Loadings3.4 Water Level (Flood) Design3.5 Missile Protection3.6 Protection Against Dynamic Effects Associated with the Postulated Rupture of

Piping3.7 Seismic Design3.8 Seismic Category I Structures3.9 Mechanical Systems and Components3.10 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment3.11 Environmental Qualification of Safety-Related Mechanical and Electrical

Equipment3.12 Safety-Related Tunnels3.13 Secondary Containment and Divisional Separation Zones-Barrier Considerations3A Seismic Soil Structure Interaction Analysis3B Containment Hydrodynamic Loads3C Computer Programs Used in the Design and Analysis of Seismic Category I

Stuctures3D Computer Programs Used in the Design of Components, Equipment and Structures3E Guidelines for LBB Application3F Not Used3G Response of Structures to Containment Loads3H Design Details and Evaluation Results of Seismic Category I Structures3I Equipment Qualification Environmental Design Criteria3J Not Used

Table of Contents

Thi d d i h F M k 4 0 3

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Rev. 1


3K Designated NEDE-24326-1-P Material Which May Not Change Without Prior NRCStaff Approval

3L Evaluation of Postulated Ruptures in High Energy Pipes3M Resolution of Intersystem LOCA for ABWR

4.0 Reactor4.1 Summary Description4.2 Fuel System Design4.3 Nuclear Design4.4 Thermal–Hydraulic Design4.5 Reactor Materials4.6 Functional Design of Reactivity Control System4A Typical Control Rod Patterns and Associated Power Distribution for ABWR4B Fuel Licensing Acceptance Criteria4C Control Rod Licensing Acceptance Criteria4D Reference Fuel Design Compliance with Acceptance Criteria

5.0 Reactor Coolant System and Connected Systems5.1 Summary Description5.2 Integrity of Reactor Coolant Pressure Boundary5.3 Reactor Vessel5.4 Component and Subsystem Design5A Method Of Compliance For Regulatory Guide 1.1505B RHR Injection Flow And Heat Capacity Analysis Outlines

6 Engineered Safety Features6.0 General6.1 Engineered Safety Feature Materials6.2 Containment Systems6.3 Emergency Core Cooling Systems6.4 Habitability Systems6.5 Fission Products Removal and Control Systems6.6 Preservice and Inservice Inspection and Testing of Class 2 and 3 Components

and Piping6.7 High Pressure Nitrogen Gas Supply System6A Regulatory Guide 1.52, Section C, Compliance Assessment6B SRP 6.5.1, Table 6.5.1-1 Compliance Assessment6C Containment Debris Protection for ECCS Strainers6D HPCF Analysis Outlines6E Additional Bypass Leakage Considerations

7.0 Instrumentation and Control Systems7.1 Introduction7.2 Reactor Protection (Trip) System (RPS)—Instrumentation and Controls7.3 Engineered Safety Feature Systems, Instrumentation and Control7.4 Systems Required for Safe Shutdown7.5 Information Systems Important to Safety7.6 All Other Instrumentation Systems Required for Safety

Table of Contents (Continued)

Table of Contents -iii

Rev. 0


7.7 Control Systems Not Required for Safety7.8 COL License Information7A Design Response to Appendix B, ABWR LRB Instrumentation and Controls7B Implementation Requirements for Hardware/Software Development7C Defense Against Common-Mode Failure in Safety-Related, Software-Based I&C

8.0 Electric Power8.1 Introduction8.2 Offsite Power Systems8.3 Onsite Power Systems8A Miscellaneous Electrical Systems

9.0 Auxiliary Systems9.1 Fuel Storage and Handling9.2 Water Systems9.3 Process Auxiliaries9.4 Air Conditioning, Heating, Cooling and Ventilation Systems9.5 Other Auxiliary Systems9A Fire Hazard Analysis9B Summary of Analysis Supporting Fire Protection Design Requirements9C Regulatory Guide 1.52, Section C, Compliance Assessment9D SRP 6.5.1, Table 6.5.1-1 Compliance Assessment

10.0 Steam and Power Conversion System10.1 Summary Description10.2 Turbine Generator10.3 Main Steam Supply System10.4 Other Features of Steam and Power Conversion System

11.0 Radioactive Waste Management11.1 Source Terms11.2 Liquid Waste Management11.3 Gaseous Waste Management System11.4 Solid Waste Management System11.5 Process and Effluent Radiological Monitoring and Sampling Systems11.6 Offsite Radiological Monitoring Program11A Radioactive Waste Management Additional Information

12.0 Radiation Protection12.1 Ensuring that Occupational Radiation Exposures are ALARA12.2 Radiation Sources12.3 Radiation Protection Design Features12.4 Dose Assessment12.5 Health Physics Program12A Appendix 12A Calculation of Airborne Radionuclides

13.0 Conduct of Operations13.1 Organizational Structure of Applicant


iv Table of Contents

Rev. 1


13.2 Training13.3 Emergency Planning13.4 Review and Audit13.5 Plant Procedures13.6 Physical Security

14.0 Initial Test Program14.1 Specific Information to be Included in Preliminary Safety Analysis Reports14.2 Specific Information to be Included in Final Safety Analysis Reports14.3 Tier 1 Selection Criteria and Processes

15.0 Accident and Analysis15.1 Decrease in Reactor Coolant Temperature15.2 Increase in Reactor Pressure15.3 Decrease in Reactor Coolant System Flow Rate15.4 Reactivity and Power Distribution Anomalies15.5 Increase in Reactor Coolant Inventory15.6 Decrease in Reactor Coolant Inventory15.7 Radioactive Release from Subsystems and Components15.8 Anticipated Transients Without Scram15A Plant Nuclear Safety Operational Analysis (NSOA)15B Failure Modes and Effects Analysis (FMEA)15C Not Used15D Probability Analysis of Pressure Regulator Downscale Failure15E ATWS Performance Evaluation15F LOCA Inventory Curves

16.0 Technical Specifications1.0 Use and Application2.0 Safety Limits (SLs)3.0 Limiting Condition for Operation (LCO) Applicability3.0 Surviellance Requirement (SR) Applicability4.0 Design Features5.0 Administrative Controls

17 Quality Assurance17.0 Introduction17.1 Quality Assurance During Design and Construction17.2 Quality Assurance During the Operations Phase17.3 Reliability Assurance Program During Design Phase

18.0 Human Factors Engineering18.1 Introduction18.2 Design Goals and Design Bases18.3 Planning, Development, and Design18.4 Control Room Standard Design Features18.5 Remote Shutdown System18.6 Systems Integration


Table of Contents -v

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18.7 Detailed Design of the Operator Interface System18.8 COL License Information18A Emergency Procedure Guidelines18B Differences Between BWROG EPG Revision 4 and ABWR EPG18C Operator Interface Equipment Characterization18D Emergency Procedures Guidelines—Input Data and Calculation Results18E ABWR Human-System Interface Design Implementation Process18F Emergency Operation Information and Controls18G Design Development and Validation Testing18H Supporting Analysis for Emergency Operation Information and Controls

19.0 Response to Severe Accident Policy Statement19.1 Purpose and Summary19.2 Introduction19.3 Internal Event Analysis19.4 External Event Analysis and Shutdown Risk Analysis19.5 Source Term Sensitivity Studies19.6 Measurement Against Goals19.7 PRA as a Design Tool19.8 Important Features Identified by the ABWR PRA19.9 COL License Information19.10 Assumptions and Insights Related to Systems Outside of ABWR Design

Certification19.11 Human Action Overview19.12 Input to the Reliability Assurance Program19.13 Summary of Insights Gained from the PRA19A Response to CP/ML Rule 10CFR50.34(f)19B Resolution of USIs and GSIs 19C Design Considerations Reducing Sabotage Risk19D Probabilistic Evaluations19E Deterministic Evaluations19EA Direct Containment Heating19EB Fuel Coolant Interactions19EC Debris Coolability and Core Concrete Interaction19ED Corium Shield19EE Suppression Pool Bypass19F Containment Ultimate Strength19FA Containment Ultimate Strength19G Not Used19H Seismic Capacity Analysis19I Seismic Margins Analysis19J Not Used19K PRA-Based Reliability and Maintenance19L ABWR Shutdown Risk Evaluation19M Fire Protection Probabilistic Risk Assessment19N Analysis of Common-Cause Failure of Multiplex Equipment19O Not Used19P Evaluation of Potential Modifications to the ABWR Design


-vi Table of Contents

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19Q ABWR Shutdown Risk Assessment19QA Fault Trees19QB DHR Reliability Study19QC Review of Significant Shutdown Events: Electrical Power and Decay Heat Removal19R Probabilistic Flooding Analysis

20.0 Question and Response Guide20.1 Question Index20.2 Questions20.3 Questions/Responses20.4 References20A ODYNA/REDYA20B Equipment Data Base

21.0 Engineering Drawings


List of Acronyms vii

Rev. 0


AAC Alternate AC

ABS Absolute

ABWR Advanced Boiling Water Reactor

AC Alternating Current

ACI American Concrete Institute

ACU Air Conditioning Unit

ACWIA AC - Independent Water Addition

ACRS Advisory Committee on Reactor Safety

ACS Atmospheric Control System

ADS Automatic Depressurization System

AEOD Office of Analysis and Evaluation of Operational Data

AFIP Automatic Fixed In-Core Probe

AFPC Augmented Fuel Pool Cooling

AFW Auxiliary Feedwater

AISC American Institute of Steel Construction

ALARA As Low As Reasonably Achievable

ALF Automated Load Following

AMB Ambient

AMG Accident Management Guidelines

ANI Basic Fire Protection for Nuclear Power Plants

ANS American Nuclear Society

ANSI American National Standards Institute

AOO Anticipated Operational Occurrences

API American Petroleum Institute

APR Automatic Power Regulator

APRM Average Power Range Monitor

APRS Automatic Power Regulator System

ARD Anti-Rotation Device

ARI Alternate Rod Insertion

ARMC Automated Rod Movement Control

ARM Area Radiation Monitoring System

ARS Acceleration Response Spectrum

ASD Adjustable Speed Drive

ASHRAE American Society of Heating, Refrigerating and Air-Conditioning Engineers, Inc.

ASI Adverse System Interactions

ASL Assumed System Loss

List of Acronyms


viii List of Acronyms

Rev. 0


ASME American Society of Mechanical Engineers

ASTM American Society for Testing and Materials

ATIP Automatic Traversing Incore Probe

ATLM Automated Thermal Limit Monitor

ATWS Anticipated Transient Without Scram

AWS American Welding Society

AWWA American Water Works Association

B&PV Boiler and Pressure Vessel (ASME Code)

BAS Breathing Air System

BLDG Building

BOC Bottom of Core or Beginning Cycle

BOP Balance of Plant

BPU Bypass Unit

BPWS Banked Position Withdrawal Sequence

BTP Branch Technical Position

BWR Boiling Water Reactor

BWROG BWR Owners’ Group

BWRT Backwash Receiving Tank

C/B Control Building

C/C Cooling Coil

CACS Containment Atmospheric Cleanup System

CAM Containment Atmospheric Monitoring System

CAP Cargo Access Portal

CAS Central Alarm Station

CAV Cumulative Absolute Velocity

CCDF Complimentary Cumulative Distibution Failure

CCI Core-Concrete Interaction

CCF Common Cause Failure

CCFP Conditional Containment Failure Probability

CCS Condensate Cleanup System

CCS Containment Cooling System

CCW Closed Cooling Water

CDF Core Damage Frequency

CDRL Core Damage Radiation Level

CERT Constant Extension Rate Test

CET Containment Event Tree

List of Acronyms (Continued)

List of Acronyms ix

Rev. 0


CF&CAE Condensate Feedwater and Condensate Air Extraction

CFM Core Flow Measurement

CFR Code of Federal Regulation

CFS Condensate and Feedwater System

CGCS Combustible Gas Control System

CH Chugging

CHRA Control Room Habitability Area

CHRS Containment Heat Removal System

CID Control Interface Diagram

CIS Containment Isolation System

CIV Combined Intermediate Valve or Containment Isolation Valve

CLOC Closed Loop Outside Containment

CMM Control Room Multiplexing Unit

CMP Configuration Management Plan

CMPF Common Mode Probabilistic Failure

CMU Control Room Multiplexing Unit

CO Condensation Oscillation

COL Combined Operating License

COPS Containment Overpressure Protection System

CP Construction Permit

CPDP Core Plate Differential Pressure

CP/ML Construction Permit/Manufacturing License

CPR Critical Power Ratio

CPS Condensate Purification System

CPU Central Processing Unit

CRC Cyclic Redundancy Checking

CRD Control Rod Drive

CRDH Control Rod Drive Hydraulic

CRGT Control Rod Guide Tube

CRT Cathode Ray Tube

CS Containment Spray

CS Control Switch

CST Condensate Storage Tank

CTG Combustion Turbine Generator

CUW Reactor Water Cleanup System

CV Control Valve


x List of Acronyms

Rev. 0


CVCF Constant Voltage Constant Frequency

CWS Circulating Water System

D-RAP Design Reliability Assurance Program

D/F Diaphragm Floor

D/S Dryer/Separator

DAC Design Acceptance Criteria

DAW Dry Active Waste

DBA Design Basis Accident

DBE Design Basis Event

DC Design Certification or Direct Current

DCH Direct Containment Heating

DCS Drywell Cooling System

DCV Drywell Connecting Vent

DEGB Double-Ended Guillotine Break

DEPSS Drywell Equipment and Pipe Support Structure

D/G Diesel Generator

DG Diesel Generator

DIV Division

DMC Digital Measurement and Control

DOD United States Department of Defense

DOE United States Department of Energy

DOF Degree of Freedom

DOI Dedicated Operator Interface

DOP Dioctyl Phthalate Smoke

DQR Dynamic Qualification Report

DRF Design Record File

DTM Digital Trip Module

DTS Drain Transfer System

DW Drywell

DWC Drywell Cooling

DWM Demineralized Water Makeup

E/B Electrical Building

E/C Erosion/Corrosion

EAB Exclusion Area Boundary

EBVS Electrical Building Ventilation System

ECCS Emergency Core Cooling System


List of Acronyms xi

Rev. 0


ECLL Electric Room Combustible Loading Limit

ECP Electrochemical Potential or Engineering Computer Program

EDG Emergency Diesel Generator

EDM Electrodischarge Machining

EHC Electrohydraulic Control

EMC Electromagnetic Compatibility

EMI Electromagnetic Interference

EMS Essential Multiplexing System

EOEC End of Equilibrium Cycle

EOF Emergency Operations Facility

EPD Electric Power Distribution

EPFM Elastic-Plastic Fracture Mechanics

EPG Emergency Procedure Guideline

EPRI Electrical Power Research Institute

EPZ Emergency Planning Zone

EQD Enviromental Qualification Document

ESD Electrostatic Discharge

ESF Engineered Safety Feature

ESFAS Engineered Safety Features Actuation System

ESS Extraction Steam System

ESW Essential Service Water

ETA Event Tree Analysis

ETS Emergency Trip System

F/D Filter-Demineralizer

FATT Fracture Appearance Transition Temperature

FCS Feedwater Control System

FCS Flammability Control System

FCU Fan Coil Unit

FCV Flow Control Valve

FDA Final Design Approval

FDSA Filled Drum Stock Area

FDWC Feedwater Control

FHA Fuel Handling Accident

FHB Fuel Handling Building

FIV Flow-Induced Vibration

FIVE Fire Induced Vulnerability Evaluation


xii List of Acronyms

Rev. 0


FMCRD Fine Motion Control Rod Drive

FMDC Fine Motion Driver Cabinet

FMEA Failure Mode and Effects Analysis

FN Ferrite Number

FPC Fuel Pool Cooling and Cleanup

FPS Fire Protection System

FPS Freeze Protection System

FTDC Fault-Tolerant Digital Controller

FW Feedwater

FWHD Feedwater Heater and Drain System

FWLB Feedwater Line Break

GCS Generator Cooling System

GDC General Design Criterion

GE General Electric

GEN Generator

GERIS GE Reactor Vessel Inspection System

GESSAR General Electric Standard Safety Analysis Report

GETAB General Electric Thermal Analysis Basis

GL Grade Level or Generic Letter

GND Ground

GSC Gland Seal Condenser

GSOS Generator Sealing Oil System

GTO Gate-Turn-Off

GWM Gaseous Waste Management

HAZ Heat-Affected Zone

HBS House Boiler System

HCLPF High Confidence Low Probability of Failure

HCSR Heating Steam Condensate Receiver

HCU Hydraulic Control Unit

HCW High Conductivity Waste

HECW HVAC Emergency Cooling Water

HELB High-Energy Line Break

HELSA High-Energy Line-Separation Analysis

HEM Homogeneous Equilibrium Model

HEP Human Error Probability

HEPA High Energy Particulate Air


List of Acronyms xiii

Rev. 0


HFE Human Factors Engineering

HFT Hot Functional Test

HGCS Hydrogen Gas Cooling System

HI Hydrogen Iodide

HIC High Integrity Containers

HNCW HVAC Normal Cooling Water

HP High Pressure

HPCF High Pressure Core Flooder

HPCI High Pressure Core Injection

HPME High Pressure Melt Ejection

HPIN High Pressure Nitrogen Gas Supply

HSCWRS Heating Steam and Condensate Water Return System

HSD Hot Shower Drain

HSI Human-System Interfaces

HSSS Hardware/Software System Specification

HTF High Temperature Failure

HTO Tritiated Oxide

HVAC Heating, Ventilating, and Air Conditioning

HVG High Valve Gate

HVT Horizontal Vent Test

HWC Hydrogen Water Chemistry

HWHS Hot Water Heating System

HX Heat Exchanger

I&C Instrumentation and Control

IAS Instrument Air System

IASCC Irradiation Assisted Stress Corrosion Cracking

IBA Intermediate Break Accident

IBD Interlocking Block Diagram

ICC Inadequate Core Cooling

ICD Interface Control Diagram

ICEA Insulated Cable Engineer Association

ICGT In-Core Guide Tube

ICM Incore Monitoring

ICS Integrated Control System

IDCOR Industry Degraded Core Rulemaking

IE Inspection and Enforcement


xiv List of Acronyms

Rev. 0


IED Instrument Electrical Diagram

IEEE Institute of Electrical and Electronics Engineers

IGSCC Intergranular Stress Corrosion Cracking

ILRT Integrated Leak Rate Test

IN Information Notice

INPO Institute of Nuclear Power Operations

INST Instrumentation

IORV Inadvertently Open Relief Valves

IOT Infrequent Operational Transients

ISA Instrument Society of America

ISI In-Service Inspection

ISLOCA Intersystem Loss-of-Coolant Accident

ISMA Independent Support Motion Response Spectrum Analysis

IST Inservice Testing

ITAAC Inspection, Tests, Analyses, and Acceptance Criteria

ITP Initial Test Program

KAG Key Assumptions and Groundrules

L/D Lower Drywell

LBB Leak-Before-Break

LBHS Large Bore Hydraulic Snubber

LCO Limiting Condition for Operation

LCP Local Control Panels

LCW Low Conductivity Waste

LD Load Driver

LDF Lower Drywell Flooder

LDS Leak Detection and Isolation System

LDW Lower Drywell

LDWI Lower Drywell Injection

LER Licensing Event Report

LERE Licensing Event Report Evaluation

LFCV Low Flow Control Valve

LOCA Loss-of-Coolant Accident

LOOP Loss of Offsite Power

LOPP Loss of Preferred Power

LP Low Pressure

LPFL Low Pressure Flooder


List of Acronyms xv

Rev. 0


LPMS Loose-Parts Monitoring System

LPRM Local Power Range Monitor

LPSP Low Power Set Point

LPZ Low Population Zone

LRB Licensing Review Bases

LRMS Liquid Radwaste Management System

LSPS Lighting and Servicing Power Supply

LTA Lead Test Assemblies

LVDT Linear Variable Differential Transformers

LWL Low Water Level

LWR Light Water Reactor

M/C Metal-Clad

MAAP Modular Accident Analysis Program

MAPLHGR Maximum Average Planar Linear Heat Generation Rate

MBA Misplaced Bundle Accident

MCAE Main Control Area Envelope

MCC Motor Control Center

MCES Main Condenser Evacuation System

MCPR Minimal Critical Power Ratio

MCR Main Control Room

MCU Multiplexer Control Unit

MEB NRC Mechanical Engineering Branch

MG Motor Generator

MIL United States Military Standard

MMI Man-Machine Interface

MOFB Mis-oriented Fuel Bundle

MOV Motor-Operated Valve

MPC Maximum Permissible Concentration

MPCWLL Maximum Primary Containment Water Level Limit

MPL Master Parts List

MPT Main Power Transformer

MRBM Multi-Channel Rod Block Monitor

MS Multiplexing System

MSF Main Steam Flow

MSL Main Steamline

MSIV Main Steamline Isolation Valve


xvi List of Acronyms

Rev. 0


MSR Moisture Separator/Reheater

MSV Mean Square Voltage

MTBF Mean Time Between Failure

MTSV Main Turbine Stop Valve

MT Main Turbine

MTTR Mean Time to Repair

MUWC Makeup Water Condensate

MUWP Makeup Water Purified

MVA Million Volt Amps

MWP Makeup Water Preparation

MWS Makeup Water System

MUX Multiplexing System

NBR Nuclear Boiler Rated

NBS Nuclear Boiler System

NCLL Normal Combustible Loading Limit

NDE Nondestructive Examination

NDTT Nil Ductility Transition Temperature

NELS Non-Class 1E Emergency Lighting Subsystems

NEMA National Electrical Manufactures Association

NEMS Non-Essential Multiplexing System

NFPA National Fire Protection Association

NG Nuclear Grade

NMS Neutron Monitoring System

NPAR Nuclear Plant Aging Research

NPSH Net Positive Suction Head

NRC Nuclear Regulatory Commission

NRHX Non-Regenerative Heat Exchanger

NRR NRC Office of Nuclear Reactor Regulation

NSAC Nuclear Safety Analysis Corporation

NSD Non-Radioactive Storm Drain

NSLS Non-Class 1E Standby Lighting Subsystems

NSOA Nuclear Safety Operational Analysis

NSS Nuclear Safety Systems

NSSS Nuclear Steam Supply System

O-RAP Operational Reliability Assurance Program

ODYN One Dimensional Dynamic Model


List of Acronyms xvii

Rev. 0


OGS Off Gas System

OIS Oxygen Injection System

OLMCPR Operating Limit Minimum Critical Power Ratio

OL Operating License

OLU Output Logic Unit

OPRM Oscillating Power Range Monitor

ORAP Operational Reliabilty Assurance Program

OSC Operational Support Center

OST Oil Storage and Transfer

P/C Power Center

P&ID Piping and Instrumentation Diagram

PaA Pascal Absolute

PaG Pascal Gage

PAMS Post Accident Monitoring System

PASS Post-Accident Sampling System

PBX Private Branch Exchange

PCB Primary Containment Boundary

PCHS Power Cycle Heat Sink

PCP Process Control Program

PCS Power Conversion Systems

PCS Process Computer System

PCS Process Control Systems

PCT Peak Cladding Temperature

PCV Primary Containment Vessel

PCW Plant Chilled Water

PDC Principal Design Criteria

PDDP Pump Deck Differential Pressure

PFD Process Flow Diagram

PG Power Generation

PGA Peak Ground Acceleration

PGC Power Generation Control Subsystem

PHCS Power Cycle Heat Sink

PIP Plant Investment Protection

PMCS Performance Monitoring and Control Subsystem

PMF Probable Maximum Flood

PMG Plant Main Generator


xviii List of Acronyms

Rev. 0


POC Product of Combustion

POP Peak Overpressure

PORV Power Operated Relief Valve

PQL Product Quality Checklist

PRA Peak Recording Accelerographs

PRA Probabilistic Risk Assessment

PRDF Pressure Regulator Downscale Failure

PRM Process Radiation Monitoring

PRS Pressure Relief System

PRV Pressure Isloation Valve

PS Pipe Space

PSD Power Spectral Density

PSI Pre-Service Inspection

PSS Process Sampling System

PWR Pressurized Water Reactor

PSW Potable and Sanitary Water

QA Quality Assurance

R/B Reactor Building

RACC Rod Action Control Cabinet

RAI Request for Additional Information

RAP Reliability Assurance Program

RAPI Rod Action and Position Information

RAT Reserve Auxiliary Transformer

RBCC Rod Brake Controller Cabinet

RBCCW Reactor Building Closed Cooling Water

RBVS Reactor Building Ventilation System

RBVSRM Reactor Building Ventilation System Radiation Monitoring

RCC Remote Communication Cabinet

RCCW Reactor Component Cooling Water

RCCV Reinforced Concrete Containment Vessel

RCIC Reactor Core Isolation Cooling

RCIS Rod Control and Information System

RCM Reactor Coolant Makeup System

RCP Reactor Coolant Pump

RCPB Reactor Coolant Pressure Boundary

RCS Reactor Coolant System


List of Acronyms xix

Rev. 0


RCW Reactor Building Cooling Water

RDA Rod Drop Accident

RDTS Radioactive Drain Transfer System

RECHAR Recombiner and Ambient Temperature Charcoal Absorption

RFCS Recirculation Flow Control System

RFI Radio Frequency Interference

RFP Reactor Feedwater Pump

RG Regulatory Guide

RHR Residual Heat Removal

RHX Regenerative Heat Exchanger

RIC Reactor Island Complex

RICSIL GE Rapid Communication Service Information Letter

RIP Reactor Internal Pump

RM Recirculation Motor

RMC Recirculation Motor Cooling

RMHX Recirculation Motor Heat Exchanger

RMISS Recirculation Motor Inflatable Shaft Seal

RMP Recirculation Motor Purge

RMU Remote Multiplexing Unit

RO Reverse Osmosis

RPS Reactor Protection System

RPT Recirculation Pump Trip

RPV Reactor Pressure Vessel

RRPS Reference Rod Pull Sequence

RRS Reactor Recirculation System

RSM Rod Server Module

RSS Remote Shutdown System

RSW Reactor Service Water

RSW Reactor Shield Wall

RW/B Radwaste Building

RWCPS Radioactive Waste Control Panel System

RWE Rod Withdrawal Error

RWM Rod Worth Minimizer

RWP Radiation Work Permit

RWST Refueling Water Storage Tank

RVSS Reactor Vessel Support Structure


/xx

xx List of Acronyms

Rev. 0


S/B Service Building

S/DRSRO Single/Dual Rod Sequence Restriction Override

S/P Suppression Pool

S&PC Steam and Power Conversion

SACF Single Active Component Failure

SAM Sampling System

SAMDA Severe Accident Mitigation Design Alternatives

SAS Service Air System

SB&PC Steam Bypass and Pressure Control

SBO Station Blackout

SBWR Simplified Boiling Water Reactor

SC Shutdown Cooling

SCB Secondary Containment Boundary

SCG Startup Coordinating Group

SCF Single Component Failure

SCRAM Reactor Trip (Safety Control Rod Axe Man)

SCRRI Selected Control Rod Run-In

SD Storm Drain

SDC Safety Design Criteria or Shutdown Cooling

SECY Office of the Secretary of the Commission

SELS Class 1E Associated Emergency Lighting System

SEP Standby Electrical Power

SER Safety Evaluation Report

SGTS Standby Gas Treatment System

SIL GE Service Information Letter

SIT Structural Integrity Test

SJAE Steam Jet Air Ejector

SLCS Standby Liquid Control System

SLD Single Line Diagram

SLMCPR Safety Limit Minimum Critical Power Ratio

SLU Safety System Logic Unit

SMA Seismic Margins Analysis

SMDM Stepping Motor Driver Modules

SMP Software Management Plan

SMS Seismic Monitoring System

SOE Single Operator Error

List of Acronyms xxi

Rev. 0


SORV Stuck Open Relief Valve

SOT System Operational Transients

SPC Suppression Pool Cooling

SPCU Suppression Pool Cleanup

SPDS Safety Parameter Display System

SPTM Suppression Pool Temperature Monitoring

SR Surveillance Requirements

SREE Safety-Related Electrical Equipment

SRMS Solid Radwaste Management System

SRNM Startup Range Neutron Monitor

SROA Safety-Related Operator Action

SRP Standard Review Plan

SRSS Square-Root-of-the-Sum-of-the-Squares

SRV Safety Relief Valve

SSAR Standard Safety Analysis Report

SSAS Station Service Air System

SSC Structures, Systems and Components

SSPC Steel Structures Painting Council

SSPV Scram Solenoid Pilot Valve

SSE Safe Shutdown Earthquake

SSI Soil-Structure Interaction

SSLC Safety System Logic and Control

SSLS Class 1E Associated Standby Lighting Subsystem

SSW Station Service Water

S&PC Steam and Power Conversion

STC Surveillance Test Controller

STPT Simulated Thermal Power Trip

STR/AP Scram Time Test Recording/Analysis Panel

STS Sewage Treatment System

STTP Scram Time Test Panel

SW Switch

SWC Surge Withstand Capability

SWSA Solid Waste Storage Area

T/B Turbine Building

T-G Turbine Generator

T&M Test and Maintenance

xxii List of Acronyms

Rev. 0


TOC Top of Core

TAF Top of Active Fuel

TASS Turbine Auxiliary Steam System

TB Turbine Bypass

TBCE Turbine Building Compartment Exhaust

TBCWS Turbine Building Cooling Water System

TBE Turbine Building Exhaust

TBLOE Turbine Building Lube Oil Area Exhaust

TBS Turbine Building Supply

TBS Turbine Bypass System

TBVS Turbine Building Ventilation System

TCF Total Core Flow

TCS Turbine Control System

TCV Turbine Control Valve

TCW Turbine Building Cooling Water

TD Tornado Damper

TDH Total Developed Head

TGSS Turbine Gland Sealing System

THA Time-History Accelerographs

TIP Traversing Incore Probe or Traversing Ion Chamber

TIU Technician Interface Unit

TLU Trip Logic Unit

TMI Three Mile Island

TMSL Typical Mean Sea Level

TN Transmission Network

TRS Test Response Spectra

TS Technical Specification

TSC Technical Support Center

TSI Turbine Supervisory Instrument

TSV Turbine Stop Valve

TSW Turbine Service Water

TVAPS Time Varying Axial Power Shape

U/D Upper Drywell

UAT Unit Auxiliary Transformers

UBC Uniform Building Code

UD Upper Drywell

List of Acronyms xxiii

Rev. 0


UHS Ultimate Heat Sink

UL Underwriters Laboratory

UPS Uninterruptible Power Source

URD Utility Requirements Document

URS Ultimate Rupture Strength

USE Upper Shelf Energy

USMA Uniform Support Motion Response Spectrum Analysis

USNRC United States Nuclear Regulatory Commission

V&V Verification and Validation

VAC Volts Alternating Current

VAP Vehicle Access Portal

VDC Volts Direct Current

VDU Video Display Unit

VLC Vent Line Clearing

VWO Valves-Wide-Open

WDSC Wetwell and Drywell Spray Cooling (Mode of RHR)

WDVB Wetwell-to-Drywell Vacuum Breaker

WDVBS Wetwell-to-Drywell Vacuum Breaker System

WRL Wide Range Level

WW Wetwell

ZIS Zinc Injection System

ZSI Zone Selective Interlocks

Table of Contents 1.0-1

Rev. 1


List of Tables.......................................................................................................................... 1.0-iii

List of Figures ........................................................................................................................1.0-vii

1.0 Introduction and General Description of Plant................................................................... 1.1-1

1.1 Introduction ........................................................................................................................... 1.1-11.1.1 Format and Content............................................................................................... 1.1-11.1.2 ABWR Standard Plant Scope ................................................................................. 1.1-11.1.3 Engineering Documentation................................................................................. 1.1-21.1.4 Design Process ........................................................................................................ 1.1-31.1.5 Type of License Required ...................................................................................... 1.1-31.1.6 Number of Plant Units ........................................................................................... 1.1-31.1.7 Description of Location ......................................................................................... 1.1-31.1.8 Type of Nuclear Steam Supply .............................................................................. 1.1-31.1.9 Type of Containment ............................................................................................. 1.1-31.1.10 Core Thermal Power Levels................................................................................... 1.1-41.1.11 COL License Information...................................................................................... 1.1-4

1.2 General Plant Description ..................................................................................................... 1.2-11.2.1 Principal Design Criteria........................................................................................ 1.2-11.2.2 Plant Description.................................................................................................... 1.2-81.2.3 COL License Information.................................................................................... 1.2-40

1.3 Comparison Tables ................................................................................................................ 1.3-11.3.1 Nuclear Steam Supply System Design Characteristics.......................................... 1.3-11.3.2 Engineered Safety Features Design Characteristics ............................................. 1.3-11.3.3 Containment Design Characteristics..................................................................... 1.3-11.3.4 Structural Design Characteristics .......................................................................... 1.3-11.3.5 Instrumentation and Electrical Systems Design Characteristics.......................... 1.3-1

1.4 Identification of Agents and Contractors ............................................................................. 1.4-1

1.5 Requirements for Further Technical Information .............................................................. 1.5-1

1.6 GE Topical Reports and Other Documents ......................................................................... 1.6-1

1.7 Drawings ................................................................................................................................. 1.7-11.7.1 Piping and Instrumentation and Process Flow Drawings .................................... 1.7-11.7.2 Instrument, Control and Electrical Drawings....................................................... 1.7-11.7.3 ASME Standard Units Metric Conversion Factors................................................ 1.7-11.7.4 Metric Conversion to ASME Standard Units ........................................................ 1.7-11.7.5 Drawing Standards ................................................................................................. 1.7-11.7.6 COL License Information...................................................................................... 1.7-1

1.8 Conformance with Standard Review Plan and Applicability of Codes and Standards ...... 1.8-11.8.1 Conformance with Standard Review Plan............................................................. 1.8-1

Chapter 1

Table of Contents

1.0-2 Table of Contents

Rev. 0


1.8.2 Applicability of Codes and Standards ................................................................... 1.8-11.8.3 Applicability of Experience Information .............................................................. 1.8-11.8.4 COL License Information...................................................................................... 1.8-2

1.9 COL License Information ..................................................................................................... 1.9-1

Appendices

1A Response to TMI Related Matters ..........................................................................................1A-1

1AA Plant Shielding to Provide Access to Vital Areas and Protective Safety Equipment for Post-Accident Operation [II.B.2] ................................................................................................ 1AA-1

1B Not Used ..................................................................................................................................1B-1

1C ABWR Station Blackout Considerations ................................................................................1C-1


List of Tables 1.0-iii

Rev. 0


Table 1.3-1 Comparison of Nuclear Steam Supply System Design Characteristics................ 1.3-2

Table 1.3-2 Comparison of Engineered Safety Features Design Characteristics ................... 1.3-7

Table 1.3-3 Comparison of Containment Design Characteristics........................................... 1.3-9

Table 1.3-4 Comparison of Structural Design Characteristics .............................................. 1.3-11

Table 1.4-1 Commercial Nuclear Reactors Completed, Under Construction, or in Design by General Electric..................................................................................... 1.4-2

Table 1.6-1 Referenced Reports................................................................................................ 1.6-2

Table 1.7-1 Piping and Instrumentation and Process Flow Diagrams .................................... 1.7-2

Table 1.7-2 Instrument Engineering, Interlock Block and Single-Line Diagrams ................ 1.7-4

Table 1.7-3 Conversion to ASME Standard Units ................................................................... 1.7-6

Table 1.7-4 Conversion Tables—Metric to ASME Standard Units ......................................... 1.7-9

Table 1.7-5 Drawing Standards ............................................................................................... 1.7-11

Table 1.8-1 Summary of Differences from SRP Section 1 ....................................................... 1.8-3









Table 1.8-10 Summary of Differences from SRP Section 10 ..................................................... 1.8-7





Chapter 1

List of Tables


1.0-iv List of Tables

Rev. 0






Table 1.8-19 Standard Review Plans and Branch Technical Positions Applicable to ABWR 1.8-10

Table 1.8-20 NRC Regulatory Guides Applicable to ABWR.................................................... 1.8-25

Table 1.8-21 Industrial Codes and Standards Applicable to ABWR ....................................... 1.8-35

Table 1.8-22 Experience Information Applicable to ABWR ................................................... 1.8-48

Table 1.9-1 Summary of ABWR Standard Plant COL License Information .......................... 1.9-2

Table 1A-1 Responses to Questions Posed by Mr. C. Michelson [II.K.3(46)] ......................1A-38

Table 1AA-1 Radiation Source Comparison........................................................................... 1AA-11

Table 1AA-2 Post-Accident Emergency Core Cooling Systems and Auxiliaries ................... 1AA-12

Table 1AA-3 Post-Accident Combustible Gas Control Systems and Auxiliaries ................... 1AA-14

Table 1AA-4 Post-Accident Fission Product Removal and Control Systems and Auxiliaries ........................................................................................................... 1AA-15

Table 1AA-5 Post-Accident Instrumentation and Controls, Power and Habitability Systems and Auxiliaries ...................................................................................... 1AA-16

Table 1B-1 Not Used ..................................................................................................................1B-1

Table 1C-1 ABWR Design Compliance with 10CFR50.63 Regulations .................................1C-10

Table 1C-2 ABWR Design Compliance with Regulatory Guide 1.155...................................1C-14

Table 1C-3 ABWR Design Compliance with NUMARC 87-00 Guidelines ............................1C-30

List of Tables (Continued)

List of Figures 1.0-v

Rev. 0


Figure 1.1-1 ABWR Standard Plant Nomenclature................................................................ 1.1-5

Figure 1.1-2 Heat Balance at Rated Power ............................................................................. 1.1-6

Figure 1.2-1 Site Plan ............................................................................................................. 1.2-41

Figure 1.2-2 Reactor Building, Arrangement Elevation, Section A-A ................................. 1.2-42

Figure 1.2-2a Reactor Building, Arrangement Elevation, Section B-B ................................. 1.2-42

Figure 1.2-3 Upper Drywell, Arrangement Elevation, Section A-A ..................................... 1.2-42

Figure 1.2-3a Upper Drywell, Arrangement Elevation, Section B-B ..................................... 1.2-42

Figure 1.2-3b Lower Drywell, Arrangement Elevation, Section A-A ..................................... 1.2-42

Figure 1.2-3c Wetwell, Arrangement Elevation, Sections A-A & B-B .................................... 1.2-42

Figure 1.2-4 Reactor Building, Arrangement Plan at Elevation –8200 mm ....................... 1.2-42

Figure 1.2-5 Reactor Building, Arrangement Plan at Elevation –1700 mm ....................... 1.2-42

Figure 1.2-6 Reactor Building, Arrangement Plan at Elevation 4800/8500 mm ............... 1.2-42

Figure 1.2-7 Not Used ............................................................................................................ 1.2-42

Figure 1.2-8 Reactor Building, Arrangement Plan at Elevation 12300 mm ....................... 1.2-42




Figure 1.2-12 Reactor Building, Arrangement Plan at Elevation 31700 mm/38,200 mm .. 1.2-42

Figure 1.2-13a Drywell, Arrangement Plan at Elevation 12300 mm ....................................... 1.2-42

Figure 1.2-13b Drywell, Arrangement Plan at Elevation 15600 mm ....................................... 1.2-42

Figure 1.2-13c Drywell, Arrangement Plan at Elevation 18100 mm ....................................... 1.2-42

Figure 1.2-13d Drywell Steel, Structure at Elevation 18100 mm ............................................. 1.2-42

Figure 1.2-13e Lower Drywell, Arrangement Plan at Elevation –6600 to – 1850 mm............ 1.2-42

Figure 1.2-13f Lower Drywell, Arrangement Plan at Elevation -1850 to 1750 mm ............... 1.2-42

Figure 1.2-13g Lower Drywell, Arrangement Plan at Elevation 1750 to 4800 mm................ 1.2-42

Figure 1.2-13h Lower Drywell, Arrangement Plan at Elevation 4800 to 6700 mm ................ 1.2-42

Chapter 1

List of Figures


1.0-vi List of Figures

Rev. 0


Figure 1.2-13i Wetwell, Arrangement Plan at Elevation – 8200 mm ..................................... 1.2-42

Figure 1.2-13j Wetwell, Arrangement Plan at Elevation – 1700 mm ..................................... 1.2-43

Figure 1.2-13k Wetwell, Arrangement Plan at Elevation 4800 mm......................................... 1.2-43

Figure 1.2-14 Control and Service Building, Arrangement Elevation, Section A-A............. 1.2-43

Figure 1.2-15 Control and Service Building, Arrangement Elevation, Section B-B............. 1.2-43

Figure 1.2-16 Control Building, Arrangement Plan at Elevation –8200 mm....................... 1.2-43

Figure 1.2-17 Control and Service Building, Arrangement Elevation –2150 mm ............... 1.2-43

Figure 1.2-18 Control and Service Building, Arrangement Elevation 3500 mm.................. 1.2-43

Figure 1.2-19 Control and Service Building, Arrangement Elevation 7900 mm.................. 1.2-43

Figure 1.2-20 Control and Service Building, Arrangement Elevation 12300 mm................ 1.2-43



Figure 1.2-23a Radwaste Building at Elevation—1500 mm..................................................... 1.2-43

Figure 1.2-23b Radwaste Building at Elevation 4800 mm........................................................ 1.2-43

Figure 1.2-23c Radwaste Building at Elevation 12300 mm...................................................... 1.2-43

Figure 1.2-23d Radwaste Building at Elevation 21000 mm...................................................... 1.2-43

Figure 1.2-23e Radwaste Building, Section A-A........................................................................ 1.2-43

Figure 1.2-23f Not Used ............................................................................................................ 1.2-43

Figure 1.2-23g Not Used ............................................................................................................ 1.2-43

Figure 1.2-24 Turbine Building, General Arrangement at Elevation 5300 mm................... 1.2-43

Figure 1.2-25 Turbine Building, General Arrangement at Elevation 12300 mm................. 1.2-43



Figure 1.2-28 Turbine Building, General Arrangement, Longitudinal Section A-A ........... 1.2-43

Figure 1.2-29 Turbine Building, General Arrangement, Section B-B.................................. 1.2-44

Figure 1.2-30 Turbine Building, General Arrangement, Section C-C ................................. 1.2-44

List of Figures (Continued)

List of Figures 1.0-vii

Rev. 0


Figure 1.2-31 Turbine Building, General Arrangement, Section D-D................................. 1.2-44

Figure 1.7-1 Piping and Instrumentation Diagram Symbols (Sheets 1–2) ......................... 1.7-21

Figure 1.7-2 Graphical Symbols for Use in IBDs .................................................................. 1.7-22

Figure 1.7-3 Graphical Symbols for Use in Electrical SLDs................................................. 1.7-30

List of Figures (Continued)

/viii

Introduction 1.1-1

Rev. 0


1.0 Introduction and General Description of Plant

1.1 Introduction

1.1.1 Format and Content

Tier 2 is written in accordance with Regulatory Guide (RG) 1.70. For consistency with NUREG-0800, Tier 2 includes Section 15.8, which addresses anticipated transients without scram (ATWS), and Chapter 18, which addresses human factors. In addition, GE’s response to TMI related matters is presented in Appendix 1A. Appendix 1C describes the ABWR station blackout considerations.

GE’s response to the severe accident policy statement is provided in Chapter 19 of Tier 2. Chapter 20 is included to provide a Question and Response guide. Chapter 21 provides the engineering drawings.

1.1.2 ABWR Standard Plant Scope

The ABWR Standard Plant includes all buildings which are dedicated exclusively or primarily to housing systems and the equipment related to the nuclear system or controls access to this equipment and systems. There are five such buildings within the scope of the ABWR Standard Plant:

(1) Reactor Building (including containment)

(2) Service Building

(3) Control Building

(4) Turbine Building

(5) Radwaste Building

In addition to these buildings and their contents, the ABWR Standard Plant provides the supporting facilities shown in Figure 1.2-1. A detailed listing of structures and systems for the ABWR Standard Plant scope of design is provided in Table 3.2-1.

The ABWR evolutionary design provides an essentially complete nuclear power plant except for site-specific elements. The site-specific elements are included as representative conceptual designs with interface requirements sufficient for the final safety analysis and design-specific probabilistic risk assessment in accordance with 10CFR52.47(a) (1) (vii) and (b) (1). Unless otherwise noted, the following site-specific elements are outside the scope of the ABWR Standard Plant:

(1) Ultimate heat sink (9.2.5), interfaces with reactor service water (spray pond, conceptual)

1.1-2 Introduction

Rev. 0


(2) Offsite power (8.2.4), transmission (The offsite power transmission network is out of scope starting from the low voltage terminals of the main and reserve transformers, reference conceptual design is provided.)

(3) Makeup water (9.2.8), preparation (well and treatment facilities, conceptual)

(4) Potable and sanitary water systems (9.2.4), partial (Portions inside the buildings of Figure 1.2-1 are in scope. All other portions are conceptual and outside scope of the standard ABWR design.)

(5) Reactor service water (9.2.15), rejects heat to the ultimate heat sink, partial (Portions inside the buildings of Figure 1.2-1 are in scope. All other portions are conceptual [pumps, valves, pipes, strainers and other equipment (Figure 9.2-7)] and are outside the scope of the standard ABWR design.)

(6) Turbine service water (9.2.16), rejects heat to the power cycle heat sink, partial (Portions inside the buildings of Figure 1.2-1 are in scope. All other portions are conceptual [pumps, valves, pipes, strainers and other equipment (Figure 9.2-8)] and are outside the scope of the standard ABWR design.)

(7) Communications (9.5.2), partial (Communication equipment inside the buildings of Figure 1.2-1 are in scope. All other portions, including connections to offsite networks are outside the scope of the standard ABWR design.)

(8) Site security (13.6.2)

(9) Circulating water system (10.4.5), circulates water (Portions inside the buildings of Figure 1.2-1 are in scope. All other portions are conceptual [pumps, valves, pipes, strainers and other equipment (Figure 10.4-3)] and are outside the scope of the standard ABWR design. This system includes the power cycle heat sink which provides a heat sink for the Circulating Water and Turbine Service Water Systems-cooling tower with makeup water and chemical control, conceptual

(10) Heating, ventilating and air conditioning (9.4), partial (Involving potential need for toxic gas monitors).

A detailed listing of the above site-specific elements is also provided in Table 3.2-1.

1.1.3 Engineering Documentation

Engineering documentation for the ABWR Standard Plant is listed on Master Parts List

Introduction 1.1-3

Rev. 1


(MPL) No. 18NS07A03*. This MPL is a controlled list, structured by system, which contains the identification of hardware and software documentation that defines the ABWR Standard Plant.

1.1.4 Design Process

GE and its associates control the review and approval of ABWR Common Engineering design documents with a procedure using the Engineering Review Memorandum (ERM). Evidence of design verification is entered into the design records of the responsible design organization. For engineering documents prepared uniquely by GE for the U.S. ABWR, changes to engineering documents are entered into the GE design record files. A COL applicant will establish the design, including the supporting detailed design documentation, consistent with the design control document referenced in the certified design rule. See Subsection 1.1.11.1 for COL license information requirements.

1.1.5 Type of License Required

Tier 2 is submitted in support of the application for design certification (DC) for the ABWR Standard Plant.

1.1.6 Number of Plant Units

For the purpose of this document, only a single standard plant will be considered.

1.1.7 Description of Location

This plant can be constructed at any location which meets the parameters identified in Chapter 2.

1.1.8 Type of Nuclear Steam Supply

This plant will have a boiling water reactor (BWR) nuclear steam supply system (NSSS) designed and supplied by GE and designated as ABWR.

1.1.9 Type of Containment

The ABWR will have a low-leakage containment vessel which comprises the drywell and pressure suppression chamber. The containment vessel is a cylindrical steel-lined reinforced concrete structure integrated with the Reactor Building. The containment nomenclature is specified in Figure 1.1-1.

* GE Proprietary

1.1-4 Introduction

Rev. 0


1.1.10 Core Thermal Power Levels

The information presented in Tier 2 pertains to one reactor unit with a rated power level of 3926 MWt and a design power level of 4005 MWt. The station utilizes a single-cycle, forced-circulation BWR. The heat balance for rated power is shown in Figure 1.1-2. The station is designed to operate at a gross electrical power output of approximately 1356 MWe and net electrical power output of approximately 1300 MWe.

1.1.11 COL License Information

1.1.11.1 Design Process to Establish Detailed Design Documentation

The COL applicant will provide the design process required to establish the detailed design documentation (see Subsection 1.1.4).

Introduction 1.1-5

Rev. 0


Figure 1.1-1 ABWR Standard Plant Nomenclature

SECONDARYCONTAINMENTBOUNDARY

DRYWELLCONNECTINGVENT

DRYWELLHEAD

PRIMARYCONTAINMENTVESSEL

CLEANZONE

UPPERDRYWELL

DIAPHRAGMFLOOR

ACCESS

HORIZ.VENT

ACCESS TUNNEL

LOWERDRYWELL

SUPPRESS.CHAMBERAIRSPACE(WETWELL)

SUPPRESS.POOL

BASE MAT

PRIMARYCONTAINMENTBOUNDARY

Note: This drawing is to illustrate thescope and requirements of thetext and is not intended to showthe final detail of the design.

1.1-6 Introduction

Rev. 0


Figure 1.1-2 Heat Balance at Rated Power

MAIN STEAM FLOW764.1 x 104 G

(TURBINE INLET)6791 P2770 h0.4 M

FEEDWATER FLOW

925.6 h215.7 T

924.8 h215.6 T

3926 MWt

TOTALCORE FLOW

52.2 x 106 G

1227 h278.3 T

203.8 h48.4 T

CLEANUPDEMINERALIZERSYSTEM

P = Pressure, kPaG = Flow, kg/hh = Enthalpy, J/gT = Temperature,°CM = % Moisture

= Isolation Valves

LEGEND Assumed System Losses

Thermal 1.1 MW7171 P2770 h0.1 M

1227 h278.3 T

966.3 h224.6 T

777.7 x 104 G 762.4 x 104 G

0.35 Δhpump

15.25 x 104 G1.64 x 104G

RIP ANDCONTROLROD DRIVEPURGE FLOW

General Plant Description 1.2-1

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1.2 General Plant Description

1.2.1 Principal Design Criteria

The principal design criteria are presented in two ways. First, they are classified as either a power generation function or a safety function. Second, they are grouped according to system. Although the distinctions between power generation or safety functions are not always clear-cut and are sometimes overlapping, the functional classification facilitates safety analyses, while the grouping by system facilitates the understanding of both the system function and design.

1.2.1.1 General Design Criteria

1.2.1.1.1 Power Generation Design Criteria

(1) The plant is designed to produce steam for direct use in a turbine-generator unit.

(2) Heat removal systems are provided with sufficient capacity and operational adequacy to remove heat generated in the reactor core for the full range of normal operational conditions and abnormal operational transients.

(3) Backup heat removal systems are provided to remove decay heat generated in the core under circumstances wherein the normal operational heat removal systems become inoperative. The capacity of such systems is adequate to prevent fuel cladding damage.

(4) The fuel cladding, in conjunction with other plant systems, is designed to retain integrity so that the consequences of any failures are within acceptable limits throughout the range of normal operational conditions and abnormal operational transients for the design life of the fuel.

(5) Control equipment is provided to allow the reactor to respond automatically to load changes and abnormal operational transients.

(6) Reactor power level is manually controllable.

(7) Control of the reactor is possible from a single location.

(8) Reactor controls, including alarms, are arranged to allow the operator to rapidly assess the condition of the reactor system and locate system malfunctions.

(9) Interlocks or other automatic equipment are provided as backup to procedural control to avoid conditions requiring the functioning of nuclear safety systems or engineered safety features.

1.2-2 General Plant Description

Rev. 0


(10) The station is designed for routine continuous operation whereby steam activation products, fission products, corrosion products, and coolant dissociation products are processed to remain within acceptable limits.

1.2.1.1.2 Safety Design Criteria

(1) The station design conforms to applicable codes and standards as described in Subsection 1.8.2.

(2) The station is designed, fabricated, erected, and operated in such a way that the release of radioactive material to the environment does not exceed the limits and guideline values of applicable government regulations pertaining to the release of radioactive materials for normal operations, abnormal transients, and accidents.

(3) The reactor core is designed so its nuclear characteristics do not contribute to a divergent power transient.

(4) The reactor is designed so there is no tendency for divergent oscillation of any operating characteristic considering the interaction of the reactor with other appropriate plant systems.

(5) The design provides means by which plant operators are alerted when limits on the release of radioactive material are approached.

(6) Sufficient indications are provided to allow determination that the reactor is operating within the envelope of conditions considered safe by plant analysis.

(7) Radiation shielding is provided and access control patterns are established to allow a properly trained operating staff to control radiation doses within the limits of applicable regulations in any mode of normal plant operations.

(8) Those portions of the nuclear system that form part of the reactor coolant pressure boundary (RCPB) are designed to retain integrity as a radioactive material containment barrier following abnormal operational transients and accidents.

(9) Nuclear safety systems and engineered safety features function to assure that no damage to the RCPB results from internal pressures caused by abnormal operational transients and accidents.

(10) Where positive, precise action is immediately required in response to abnormal operational transients and accidents, such action is automatic and requires no decision or manipulation of controls by plant operations personnel.


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(11) Safety-related actions are provided by equipment of sufficient redundance and independence so that no single failure of active components, or of passive components in certain cases in the long term, will prevent the required actions.

(12) Provisions are made for control of active components of safety-related systems from the control room.

(13) Safety-related systems are designed to permit demonstration of their functional performance requirements.

(14) The design of safety-related systems, components and structures includes allowances for natural environmental disturbances such as earthquakes, floods, and storms at the station site.

(15) Standby electrical power sources have sufficient capacity to power all safety-related systems requiring electrical power concurrently.

(16) Standby electrical power sources are provided to allow prompt reactor shutdown and removal of decay heat under circumstances where normal auxiliary power is not available.

(17) A containment is provided that completely encloses the reactor systems, drywell, and suppression chambers. The containment employs the pressure suppression concept.

(18) It is possible to test primary containment integrity and leaktightness at periodic intervals.

(19) A secondary containment is provided that completely encloses the primary containment above the Reactor Building basemat. This secondary containment provides for a controlled, monitored release of any potential radioactive leakage from the primary containment.

(20) The primary containment and secondary containment, in conjunction with other safety-related features, limit radiological effects of accidents resulting in the release of radioactive material to the containment volumes to less than the prescribed acceptable limits.

(21) Provisions are made for removing energy from the primary containment as necessary to maintain the integrity of the containment system following accidents that release energy to the containment.


Rev. 0


(22) Piping that penetrates the primary containment and could serve as a path for the uncontrolled release of radioactive material to the environs is automatically isolated when necessary to limit the radiological impact from an uncontrolled release to less than acceptable limits.

(23) Emergency core cooling systems (ECCS) are provided to limit fuel cladding temperature to less than the limits of 10CFR50.46 in the event of a loss-of-coolant accident (LOCA).

(24) The ECCS provide for continuity of core cooling over the complete range of postulated break sizes in the RCPB.

(25) Operation of the ECCS is initiated automatically when required regardless of the availability of offsite power supplies and the normal generating system of the station.

(26) The control room is shielded against radiation so that continued occupancy under design basis accident conditions is possible.

(27) In the event that the control room becomes inaccessible, it is possible to bring the reactor from power range operation to cold shutdown conditions by utilizing alternative controls and equipment that are available outside the control room.

(28) Backup reactor shutdown capability independent of normal reactivity control is provided. This backup system has the capability to shut down the reactor from any normal operating condition and subsequently to maintain the shutdown condition.

(29) Fuel handling and storage facilities are designed to prevent inadvertent criticality and to maintain shielding and cooling of spent fuel as necessary to meet operating and offsite dose constraints.

(30) Systems that have redundant or backup safety functions are physically separated, and arranged so that credible events causing damage to one region of the Reactor Island complex has minimum prospect for compromising the functional capability of the redundant system.

1.2.1.2 System Criteria

The principal design criteria for particular systems are listed in the following subsections.


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1.2.1.2.1 Nuclear System Criteria

(1) The fuel cladding is a radioactive material barrier designed to retain integrity so that failures do not result in dose consequences that exceed acceptable limits throughout the design power range.

(2) The fuel cladding, in conjunction with other plant systems, is designed to retain integrity so that the consequences of any failures are within acceptable limits throughout any abnormal operational transient.

(3) Those portions of the nuclear system that form part of the RCPB are designed to retain integrity as a radioactive material barrier during normal operation and following abnormal operational transients and accidents.

(4) The capacity of the heat removal systems provided to remove heat generated in the reactor core for the full range of normal operational transients as well as for abnormal operational transients is adequate to prevent fuel cladding damage that results in dose consequences exceeding acceptable limits.

(5) The reactor is capable of being shut down automatically in sufficient time to permit decay heat removal systems to become effective following loss of operation of normal heat removal systems. The capacity of such systems is adequate to prevent fuel cladding damage.

(6) The reactor core and reactivity control system are designed such that control rod action is capable of making the core subcritical and maintaining it even with the rod of highest reactivity worth fully withdrawn and unavailable for insertion.

(7) Backup reactor shutdown capability is provided independent of normal reactivity control provisions. This backup system has the capability to shut down the reactor from any operating condition and subsequently to maintain the shutdown condition.

(8) The nuclear system is designed so there is no tendency for divergent oscillation of any operating characteristic, considering the interaction of the nuclear system with other appropriate plant systems.

1.2.1.2.2 Electrical Power Systems Criteria

Sufficient normal auxiliary and standby sources of electrical power are provided to attain prompt shutdown and continued maintenance of the station in a safe condition under all credible circumstances. The power sources are adequate to accomplish all required essential safety actions under all postulated accident conditions.


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1.2.1.2.3 Auxiliary Systems Criteria

(1) Fuel handling and storage facilities are designed to prevent inadvertent criticality and to maintain adequate shielding and cooling for spent fuel.

(2) Other auxiliary systems, such as service water, cooling water, fire protection, heating and ventilating, communications, and lighting, are designed to function as needed, during normal and/or accident conditions.

(3) Auxiliary systems that are not required to effect safe shutdown of the reactor or maintain it in a safe condition are designed so that a failure of these systems shall not prevent the essential auxiliary systems from performing their design functions.

1.2.1.2.4 Shielding and Access Control Criteria

Radiation shielding is provided and access control patterns are established to allow a properly trained operating staff to control radiation doses within the limits of applicable regulations in any normal mode of plant operation.

1.2.1.2.5 Process Control Systems Criteria

The principal design criteria for the process control systems are listed in the following subsections.

1.2.1.2.5.1 Nuclear System Process Control Criteria

(1) Control equipment is provided to allow the reactor to respond automatically to load changes within design limits.

(2) It is possible to control the reactor power level manually.

(3) Nuclear systems process displays, controls and alarms are arranged to allow the operator to rapidly assess the condition of the nuclear system and to locate process system malfunctions.

1.2.1.2.5.2 Electrical Power System Process Control Criteria

(1) The Class 1E power systems are designed with three divisions with any two divisions being adequate to safely place the unit in the hot shutdown condition.

(2) Protective relaying is used to detect and isolate faulted equipment from the system with a minimum of disturbance in the event of equipment failure.


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(3) Voltage relays are used on the emergency equipment buses to disconnect the normal source in the event of loss of offsite power and to initiate starting of the standby emergency power system diesel generators.

(4) The standby emergency power diesel generators are started and loaded automatically.

(5) Safety-related electrically operated breakers are controllable from the control room.

(6) Monitoring of essential generators, transformers, and circuits is provided in the main control room.

1.2.1.2.5.3 Power Conversion Systems Process Control Criteria

(1) Control equipment is provided to control the reactor pressure throughout its operating range.

(2) The turbine is able to respond automatically to minor changes in load.

(3) Control equipment in the feedwater system maintains the water level in the reactor vessel at the optimum level required by steam separators.

(4) Control of the power conversion equipment is possible from a central location.

1.2.1.2.6 Power Conversion Systems Criteria

Components of the power conversion systems shall be designed to perform the following basic objectives:

(1) Produce electrical power from the steam coming from the reactor, condense the steam into water, and return the water to the reactor as heated feedwater with a major portion of its gases and particulate impurities removed.

(2) Assure that any fission products or radioactivity associated with the steam and condensate during normal operation are safely contained inside the system or are released under controlled conditions in accordance with waste disposal procedures.

1.2.1.3 Plant Design and Aging Management

The COL applicant shall initiate the life cycle management program early enough in the design process to aid in the application, selection and procurement of components with optimum design life characteristics, and to develop an aging management plan capable of assuring the plant’s original design basis throughout its life.


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The aging management plan shall cover containment structures, liner plates, embedded or buried structural components, piping and components. The plan shall consider the potential causes of corrosion which ultimately may be present at the site, including the potential corrosion from copper ground mats. The plan should be initiated early in the design process so that adequate provisions for mitigation measures can be made.

In developing the life cycle management program, the COL applicant shall consider the design life requirements prescribed in the EPRI Utility Requirements Document (URD) and the insights gained from the USNRC Nuclear Plant Aging Research (NPAR) Program. (e.g. NUREG/CRs - 4731 and - 5314)

See Subsection 1.2.3.1 for COL license information.

1.2.2 Plant Description

1.2.2.1 Site Characteristics

1.2.2.1.1 Site Location

The plant is located on a site adjacent or close to a body of water with sufficient capacity for either once-through or recirculated cooling or a combination of both methods.

1.2.2.1.2 Description of Plant Environs

1.2.2.1.2.1 Meteorology

The safety-related structures and equipment are designed to retain required functions for the loads resulting from any tornado with characteristics not exceeding the values provided in Table 2.0-1.

Tornado missiles are discussed in Section 3.5.

1.2.2.1.2.2 Hydrology

The safety design basis of the plant provides that structures of safety significance will be unaffected by the hydrologic parameter envelope defined in Chapter 2.

1.2.2.1.2.3 Geology and Seismology

The structures of safety significance for the plant are designed to withstand a safe shutdown earthquake (SSE) which results in a freefield peak acceleration of 0.3g.

1.2.2.1.2.4 Shielding

Shielding is provided throughout the plant, as required, to maintain radiation levels to operating personnel and to general public within the applicable limits set forth in


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10CFR20 and 10CFR100. It is also designed to protect certain plant components from radiation exposure resulting in unacceptable alterations of material properties or activation.

1.2.2.1.3 Site Arrangements

The containment and building arrangements, including equipment locations, are shown in Figures 1.2-2 through 1.2-31. The arrangement of these structures on the plant site is shown in Figure 1.2-1.

1.2.2.2 Nuclear Steam Supply Systems

The Nuclear Steam Supply System (NSSS) includes a direct-cycle forced-circulation BWR that produces steam for direct use in the steam turbine. A heat balance showing the major parameters of the NSSS for the rated power conditions is shown in Figure 1.1-2.

1.2.2.2.1 Reactor Pressure Vessel System

The Reactor Pressure Vessel (RPV) System contains the reactor pressure vessel with the reactor internal pump (RIP) casings; core and supporting structures; the steam separators and dryers; the control rod guide tubes; the spargers for the feedwater, RHR and core flooder system; the control rod drive (CRD) housings; the incore instrumentation guide tubes and housings; and other components. The main connections to the vessel include steamlines, feedwater lines, RIPs, CRDs and incore nuclear instrument detectors, core flooder lines, RHR lines, head spray and vent lines, core plate differential pressure lines, internal pump differential pressure lines, and water level instrumentation.

A venturi-type flow restrictor is a part of the RPV nozzle configuration for each steamline. These restrictors limit the flow of steam from the reactor vessel before the main steamline isolation valves (MSIVs) are closed in case of a main steamline break outside the containment.

The RPV System provides guidance and support for the CRDs. It also distributes boron (sodium pentaborate) solution when injected from the Standby Liquid Control (SLC) System.

The RPV System restrains the CRD to prevent ejection of the control rod connected with the CRD in the event of a failure of the RCPB associated with the CRD housing weld.

CRD blowout restraints are located internal to the reactor vessel and the control rod drive. A restraint system is also provided for each RIP in order to prevent the RIP from


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becoming a missile in the event of a failure of the RCPB associated with the RIP casing weld.

The reactor vessel is designed and fabricated in accordance with applicable codes for a pressure of 8620 kPaG. The nominal operating pressure in the steam space above the separators is 7170 kPaA. The vessel is fabricated of low alloy steel and is clad internally with stainless steel or Ni-Cr-Fe Alloy (except for the top head, RIP motor casing, nozzles other than the steam outlet nozzle, and nozzle weld zones which are unclad).

The reactor core is cooled by demineralized water that enters the lower portion of the core and boils as it flows upward around the fuel rods. The steam leaving the core is dried by steam separators and dryers located in the upper portion of the reactor vessel. The steam is then directed to the turbine through the main steamlines. Each steamline is provided with two isolation valves in series, one on each side of the containment barrier.

1.2.2.2.2 Nuclear Boiler System

1.2.2.2.2.1 Main Steamline Isolation Valves

All pipelines that both penetrate the containment and offer a potential release path for radioactive material are provided with redundant isolation capabilities. Isolation valves are provided in each main steamline to isolate primary containment upon receiving an automatic or manual closure signal. Each is powered by both pneumatic pressure and spring force. These valves fulfill the following objectives:

(1) Prevent excessive damage to the fuel barrier by limiting the loss of reactor coolant from the reactor vessel resulting from either a major leak from the steam piping outside the containment or a malfunction of the pressure control system resulting in excessive steam flow from the reactor vessel.

(2) Limit the release of radioactive materials by isolating the RCPB in case of the detection of high steamline radiation.

1.2.2.2.2.2 Main Steamline Flow Instrumentation

The steam flow instrumentation is connected to the venturi type steam nozzle of the RPV. The instrumentation provides high nozzle flow isolation signals in case of a main steamline break, flow signals for the feedwater flow control system and indication in the control room.

1.2.2.2.2.3 Nuclear System Pressure Relief System

A pressure relief system consisting of safety/relief valves (SRVs) mounted on the main steamlines is provided to prevent excessive pressure inside the nuclear system as a result of operational transients or accidents.


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1.2.2.2.2.4 Automatic Depressurization System

The ADS rapidly reduces reactor vessel pressure in a loss-of-coolant accident, enabling the low-pressure RHR to deliver cooling water to the reactor vessel.

The ADS uses some of the SRVs that are part of the nuclear system pressure relief system. The SRVs used for ADS are set to open on detection of appropriate low reactor water level and high drywell pressure signals. The ADS will not be activated unless either an HPCF or RHR/low-pressure flooder loop pump is operating. This is to ensure that adequate coolant will be available to maintain reactor water level after depressurization.

1.2.2.2.2.5 Reactor Vessel Instrumentation

In addition to instrumentation for the nuclear safety systems and engineered safety features, instrumentation is provided to monitor and transmit information that can be used to assess conditions existing inside the reactor vessel and the physical condition of the vessel itself. This instrumentation monitors reactor vessel pressure, water level, coolant temperature, reactor core differential pressure, coolant flow rates, and reactor vessel head inner seal ring leakage.

1.2.2.2.3 Reactor Recirculation System

The reactor internal pumps (RIPs) are internal pumps which provide a continuous internal circulation path for the core coolant flow. The RIPs are located at the bottom of the vessel. The pump motors are enclosed in casings which are a part of the vessel. A break in the casing as described in Subsection 15B.3.4.3 will result in a leak flow that is less than the ECCS capacity, thus allowing full core coverage. The internal pumps are a wet motor design with no shaft seals, thereby providing increased reliability, reduced maintenance requirements and decreased operational radiation exposure. The RIP has a low rotating inertia. Coupled with the solid-state adjustable speed drives, the RIP can respond quickly to load transients and operator demands.

1.2.2.3 Control and Instrument Systems

1.2.2.3.1 Rod Control and Information System

The Rod Control and Information System (RCIS) provides the means by which control rods are positioned from the control room for power control. The system operates the rod drive motors to change control rod position. For operation in the normal gang movement mode, one gang of control rods can be manipulated at a time. The system includes the logic that restricts control rod movement (rod block) under certain conditions as a backup to procedural controls.


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1.2.2.3.2 Control Rod Drive System

When scram is initiated by the RPS, the Control Rod Drive (CRD) System inserts the negative reactivity necessary to shut down the reactor. Each control rod is normally controlled by an electric motor unit. When a scram signal is received, high-pressure water stored in nitrogen charged accumulators forces the control rods into the core and the electric motor drives are signalled to drive the rods into the core. Thus, the hydraulic scram action is backed up by an electrically energized insertion of the control rods.

1.2.2.3.2.1 Control Rod Braking Mechanism

An electromechanical braking mechanism is incorporated in each control rod drive to prevent ejection of the attached control rod in the event of a hydraulic line break. This action limits the rate of reactivity insertion resulting from a rod ejection accident.

1.2.2.3.2.2 Control Rod Ejection

A nuclear excursion is prevented in case of a housing failure and thus the fuel barrier is protected because, as discussed in Subsection 1.2.2.2.1, the housing and the drive are restrained internally to the vessel to prevent the control rod ejection.

1.2.2.3.3 Feedwater Control System

The Feedwater Control System (FCS) automatically controls the flow of feedwater into the reactor pressure vessel to maintain the water within the vessel at predetermined levels. A fault-tolerant triplicated, digital controller using a conventional three-element control scheme is used to accomplish this function.

1.2.2.3.4 Standby Liquid Control System

The Standby Liquid Control System (SLCS) provides an alternate method to bring the nuclear fission reaction to subcriticality and to maintain subcriticality as the reactor cools. The system makes possible an orderly and safe shutdown in the event that not enough control rods can be inserted into the reactor core to accomplish shutdown in the normal manner. The system is sized to counteract the positive reactivity effect from rated power to the cold shutdown condition.

1.2.2.3.5 Neutron Monitoring System

The Neutron Monitoring System (NMS) consists of incore neutron detectors and out-of-core electronic monitoring equipment. The NMS provides indication of neutron flux, which can be correlated to thermal power level for the entire range of flux conditions that can exist in the core. There are fixed incore sensors, the startup range neutron monitors (SRNM), which provide level indications during reactor startup and low power operation. The local power range monitors (LPRM) and average power


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range monitors (APRM) allow assessment of local and overall flux conditions during power range operation. The automatic traversing incore probe (ATIP) system provides a means to calibrate the local power range monitors. The NMS provides inputs to the RCIS to initiate rod blocks if preset flux limits or period limits for rod block are exceeded, as well as inputs to the RPS if other limits for scram are exceeded.

Those portions of the NMS that input signals to the RPS qualify as a safety-related system. The SRNM and the APRM which monitor neutron flux via incore detectors provide scram logic inputs to the RPS to initiate a scram in time to prevent excessive fuel clad damage as a result of over-power transients. The APRM system also generates a simulated thermal power signal. Both upscale neutron flux and upscale simulated thermal power are conditions which provide scram logic signals.

1.2.2.3.6 Remote Shutdown System

In the event that the control room becomes inaccessible, the reactor can be brought from power range operation to cold shutdown conditions by use of controls and equipment that are available outside the control room.

1.2.2.3.7 Reactor Protection System

The Reactor Protection System (RPS) initiates a rapid, automatic shutdown (scram) of the reactor. It acts in time to prevent fuel cladding damage and any nuclear system process barrier damage following abnormal operational transients. The RPS overrides all operator actions and process controls and is based on a failsafe design philosophy that allows appropriate protective action even if a single failure occurs.

1.2.2.3.8 Recirculation Flow Control System

During normal power operation, the speed of the reactor internal pumps (RIP) is adjusted to control flow. Adjusting RIP speed changes the coolant flow rate through the core and thereby changes the core power level. The system can automatically adjust the reactor power output to the load demand. The solid-state adjustable speed drives (ASD) provide variable-voltage/variable-frequency electrical power to the RIP motors. In response to plant needs, the recirculation flow control system adjusts the ASD power supply output to vary RIP speed, core flow, and core power.

1.2.2.3.9 Automatic Power Regulator System

The Automatic Power Regulator System is summarized in Subsection 7.7.1.7(1).

1.2.2.3.10 Steam Bypass and Pressure Control System

A turbine bypass system is provided which passes steam directly to the main condenser under the control of the pressure regulator. Steam is bypassed to the condenser whenever the reactor steaming rate exceeds the load permitted to pass to the turbine


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generator. The turbine bypass system has the capability to shed 40% of the turbine-generator rated load without reactor trip or operation of safety/relief valves. The pressure regulation system provides main turbine control valve and bypass valve flow demands so as to maintain a nearly constant reactor pressure during normal plant operation. It also provides demands to the recirculation system to adjust power level by changing reactor recirculation flow rate.

1.2.2.3.11 Process Computer (Includes PMCS, PGCS)

Online process computers are provided to monitor and log process variables and make certain analytical computations. The performance and power generation control systems are included.

1.2.2.3.12 Refueling Platform Control Computer

The refueling platform control computer provides (1) memory of all the fuel and platform positions, (2) directions for the traversable area and traveling paths, (3) directions for the speed functions for all modes of travel, and (4) control of the fuel load. The computer controls automatic or manual refueling between fuel storage and the reactor from the remote control room.

1.2.2.3.13 CRD Removal Machine Control Computer

The CRD handling machine control computer provides automatic positioning, continuous operation and prevention of erroneous operation in the stepwise removal and installation of CRDs from the remote control room.

1.2.2.4 Radiation Monitoring Systems

1.2.2.4.1 Process Radiation Monitoring System

The process radiation monitoring system measures and controls radioactivity in process and effluent streams and activate appropriate alarms and controls.

The process radiation monitoring system measures and records radiation levels associated with selected plant process streams and effluent paths leading to the environment. All effluents from the plant which are potentially radioactive are monitored.

1.2.2.4.2 Area Radiation Monitoring System

The area radiation monitoring system alerts local occupants and the control room personnel of excessive gamma radiation levels at selected locations within the plant.


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1.2.2.4.3 Containment Atmospheric Monitoring System

The Containment Atmospheric Monitoring System (CAMS) measures, records and alarms the radiation levels and the oxygen and hydrogen concentration levels in the primary containment under post-accident conditions. It is automatically put in service upon detection of LOCA conditions.

1.2.2.5 Core Cooling System

In the event of a breach in the RCPB that results in a loss of reactor coolant, three independent divisions of ECCS are provided to maintain fuel cladding below the temperature limit as defined by 10CFR50.46. Each division contains one high pressure and one low pressure inventory makeup system.

1.2.2.5.1 Residual Heat Removal System

The Residual Heat Removal (RHR) System is a system of pumps, heat exchangers, and piping that fulfills the following functions:

(1) Removes decay and sensible heat during and after plant shutdown.

(2) Injects water into the reactor vessel following a LOCA to reflood the core in conjunction with other core cooling systems (Subsection 5.5.1).

(3) Removes heat from the containment following a LOCA to limit the increase in containment pressure. This is accomplished by cooling and recirculating the suppression pool water by containment sprays.

1.2.2.5.1.1 Low Pressure Flooder

Low pressure flooding is an operating mode of each RHR system, but is discussed here because the low pressure flooder (LPFL) mode acts in conjunction with other injection systems. LPFL uses the RHR pump loops to inject cooling water into the pressure vessel. LPFL operation provides the capability of core flooding at low vessel pressure following a LOCA in time to maintain the fuel cladding below the prescribed temperature limit.

1.2.2.5.1.2 Residual Heat Removal System Containment Cooling

The RHR System is placed in operation to: (1) limit the temperature of the water in the suppression pool and the atmospheres in the drywell and suppression chamber following a design basis LOCA; (2) control the pool temperature during normal operation of the safety/relief valves and the RCIC System; and (3) reduce the pool temperature following an isolation transient. In the containment cooling mode of operation, the RHR main system pumps take suction from the suppression pool and pump the water through the RHR heat exchangers, where cooling takes place by transferring heat to the service water. The fluid is then discharged back either to the


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suppression pool, the drywell spray header, the suppression chamber spray header, or the RPV.

1.2.2.5.1.3 Wetwell/Drywell Spray

A spray system is provided for wetwell/drywell cooling in the suppression chamber and drywell air space. The wetwell/drywell spray can be initiated manually if a high containment pressure signal is received. Each subsystem is supplied from a separate redundant RHR subsystem.

1.2.2.5.2 High Pressure Core Flooder System

High pressure core flooder (HPCF) systems are provided in two divisions to maintain an adequate coolant inventory inside the reactor vessel to limit fuel cladding temperatures in the event of breaks in the reactor coolant pressure boundary. The systems are initiated by either high pressure in the drywell or low water level in the vessel. They operate independently of all other systems over the entire range of system operating pressures. The HPCF System pump motors are powered by a diesel generator if auxiliary power is not available. The systems may also be used as a backup for the RCIC System.

1.2.2.5.3 Leak Detection and Isolation System

The leak detection and isolation system (LDS) detects and monitors leakage from the reactor coolant pressure boundary and initiates isolation of the leakage source. The system initiates isolation of the process lines that penetrate the containment by closing the appropriate inboard and outboard isolation valves. LDS monitors leakage inside and outside of the drywell and annunciates excessive leakages in the control room. The following control and isolation functions are automatically performed by LDS:

(1) Isolates the main steamlines

(2) Isolates the reactor water cleanup process lines

(3) Initiates the standby gas treatment system

(4) Isolates the Reactor Building HVAC system

(5) Isolates the containment purge and vent lines

(6) Isolates the cooling water lines in the Reactor Building

(7) Isolates the RHR shutdown cooling system lines

(8) Isolates the steamline to the RCIC turbine

(9) Isolates the suppression pool cleanup system lines


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(10) Isolates the flammability control system lines

(11) Isolates the drywell sumps drain lines

(12) Isolates the fission products monitor sampling and return lines

(13) Initiates withdrawal of the automated traversing incore probe

In addition to the above functions, LDS monitors leakage inside the drywell from the following sources and annunciates the abnormal leakage levels in the control room:

(1) Fission products releases

(2) Condensate flow from the drywell air coolers

(3) Drywell sump level changes

(4) Leakages from valve stems equipped with leak-off lines

Other leakages from the FMCRDs, the SRVs and from the reactor vessel head seal flange are monitored by their respective systems.

1.2.2.5.4 Reactor Core Isolation Cooling System

The RCIC System provides makeup water to the reactor vessel when the vessel is isolated and is also part of the emergency core cooling network. The RCIC System uses a steam-driven turbine-pump unit and operates automatically in time and with sufficient coolant flow to maintain adequate water level in the reactor vessel for events defined in Section 5.4.

One division contains the RCIC System, which consists of a steam-driven turbine which drives a pump assembly and the turbine and pump accessories. The system also includes piping, valves, and instrumentation necessary to implement several flow paths. The RCIC steam supply line branches off one of the main steamlines (leaving the RPV) and goes to the RCIC turbine with drainage provision to the main condenser. The turbine exhausts to the suppression pool with vacuum breaking protection. Makeup water is supplied from the condensate storage tank (CST) or the suppression pool with preferred source being the CST. RCIC pump discharge lines include the main discharge line to the feedwater line, a test return line to the suppression pool, a minimum flow bypass line to the suppression pool and a cooling water supply line to auxiliary equipment.

Following a reactor scram, steam generation in the reactor core continues at a reduced rate due to the core fission product delay heat. The turbine condenser and the feedwater system supply the makeup water required to maintain reactor vessel inventory.


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In the event the reactor vessel is isolated, and the feedwater supply is unavailable, relief valves are provided to automatically (or remote manually) maintain vessel pressure within desirable limits. The water level in the reactor vessel drops due to continued steam generation by decay heat. Upon reaching a predetermined low level, the RCIC System is initiated automatically. The turbine-driven pump supplies water from the suppression pool or from the CST to the reactor vessel. The turbine is driven with a portion of the decay heat steam from the reactor vessel, and exhausts to the suppression pool.

In the event of a LOCA, the RCIC System, in conjunction with the two HPCF systems, is designed to pump water into the vessel from approximately 1.0 MPaG to full operating pressure. These high pressure systems, combined with the RHR low pressure flooders and ADS, make up the ECCS network which can accommodate any single failure and still shut down the reactor (see Subsection 6.3.1.1 for a detailed description of ECCS redundancy and reliability).

During RCIC operation, the wetwell suppression pool acts as the heat sink for steam generated by reactor decay heat. This results in a rise in pool water temperature. Heat exchangers in the RHR System are used to maintain pool water temperature within acceptable limits by cooling the pool water directly.

1.2.2.6 Reactor Servicing Equipment

1.2.2.6.1 Fuel Servicing Equipment

Fuel servicing equipment is summarized in Subsection 9.1.4.2.3.

1.2.2.6.2 Miscellaneous Servicing Equipment

The servicing aids equipment includes general handling fuel pool tools such as actuating poles with various end configurations. General area underwater lights and support brackets are provided to allow the lights to be positioned over the area being serviced independent of the platform. A general-purpose, plastic viewing aid is provided to float on the water surface to provide better visibility. A portable underwater closed circuit television camera may be lowered into the reactor vessel pool and/or the fuel storage pool to assist in the inspection and/or maintenance of these areas. An underwater vacuum with submersible pump and filter for cleaning.

1.2.2.6.3 Reactor Pressure Vessel Servicing Equipment

Equipment associated with servicing the reactor pressure vessel is used when the reactor is shutdown and the reactor vessel head is being removed or installed. Tools used consist of strongbacks, nut racks, stud tensioners, protectors, wrenches, etc. Lifting tools are designed for a 60-year life and for a safety factor of 10 or better with respect to the ultimate strength of the material used.


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1.2.2.6.4 RPV Internal Servicing Equipment

The majority of internal servicing equipment was designed to be attached to the refueling platform auxiliary hoist and used when the reactor is open. A variety of equipment (e.g., grapples, guides, plugs, holders, caps, strongbacks and sampling stations) is used for internal servicing. In addition to these are the RIP handling devices for repair and/or installation. Lifting tools are designed for a safety factor of 10 or better with respect to the ultimate strength of the material used.

1.2.2.6.5 Refueling Equipment

The fuel servicing equipment includes a 1.471 MN Reactor Building crane, refueling machine, and other related tools for reactor servicing.

The Reactor Building crane handles the spent fuel cask from the transport device to the cask loading pit. The refueling machine transfers the fuel assemblies between the storage area and the reactor core. New fuel bundles are handled by the Reactor Building crane. The bundles are stored in the new fuel vault on the reactor refueling floor and are transferred from the vault to the spent fuel pool with the Reactor Building crane auxiliary hook.

The handling of the reactor head, removable internals, reactor insulation, and drywell head during refueling is accomplished using the Reactor Building crane.

1.2.2.6.5.1 Refueling Interlocks

A system of interlocks that restricts movement of refueling equipment and control rods when the reactor is in the refueling and startup modes is provided to prevent an inadvertent criticality during refueling operation. The interlocks backup procedural controls that have the same objective. The interlocks affect movement of the refueling machine, refueling machine hoists, fuel grapple, and control rods.

1.2.2.6.6 Fuel Storage Facility

New and spent fuel storage racks are designed to prevent inadvertent criticality and load buckling. Sufficient cooling and shielding are provided to prevent excessive pool heatup and personnel exposure, respectively. The design of the fuel pool provides for corrosion resistance, adherence to Seismic Category I requirements, and prevention of keff from reaching 0.95 under dry or flooded conditions.

1.2.2.6.7 Undervessel Servicing Equipment

This equipment is used for the installation and removal work associated with the fine motion control rod drive (FMCRD), RIP, incore monitoring (ICM) and so on. A handling platform provides a working surface for equipment and personnel performing work in the undervessel area. The polar platform is capable of rotating 360 degrees, and


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has an FMCRD handling trolley with full traverse capability across the vessel diameters. All equipment is designed to minimize radiation exposure, contamination of surrounding equipment and reduce the number of workers required.

1.2.2.6.8 CRD Maintenance Facility

The CRD maintenance facility is located close to the primary containment and is designed and equipped to accommodate maintenance of the FMCRD, provide decontamination of the FMCRD component, perform the acceptance tests and provide storage. The facility uses manual and/or remote operation to minimize radiation exposure to the personnel and to minimize the contamination of surrounding equipment during operation. The layout of the facility is designed so as to maximize the efficiency of the personnel, thereby minimizing the number of workers required.

1.2.2.6.9 Internal Pump Maintenance Facility

The reactor internal pump (RIP) maintenance facility is located in the Reactor Building and is designed for performing maintenance work on the motor assembly and related parts. The facility is designed for one motor assembly, including decontamination in assembled and disassembled states. The facility is equipped with all tools needed for inspection of motor parts and heat exchanger tube bundles. RIP handling tools are stored outside this area.

1.2.2.6.10 Fuel Cask Cleaning Facility

The fuel cask cleaning facility provides for empty casks to be checked for contamination and cleaned of road dirt, moved into the Reactor Building airlock, inspected for damage, and raised to the refueling floor cask pit. The closure head is removed and stored in the adjacent cask washdown pit, while the canal gates between the cask pit and spent fuel pool are removed and the spent fuel is transferred to fill the cask. The canal gates and closure head are replaced and the cask is lifted to the washdown pit. The cask is decontaminated with high pressure water sprays, chemicals and hand scrubbing to the level required for offsite transport. Smear tests are performed to verify cleaning before the filled cask is lowered to the airlock, mounted on the transport vehicle and moved out of the Reactor Building.

1.2.2.6.11 Plant Startup Test Equipment

Plant startup test equipment is a combination of strain gauges, accelerometers, temperature detectors, photo cells, pressure transducers and other associated instrumentation for conducting special startup and reactor internal vibration tests.

1.2.2.6.12 Inservice Inspection Equipment

Inservice inspection equipment are coordinated ultrasonic, eddy current and visual systems needed for incore housing, stub tube, feedwater nozzle, RPV inside and outside


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diameters (GERIS 2000), RPV internals and head studs, shroud head bolts, and piping (SMART 2000) inspections and examinations.

1.2.2.7 Reactor Auxiliary Systems

1.2.2.7.1 Reactor Water Cleanup System

The Reactor Water Cleanup System (CUW) recirculates a portion of reactor coolant through a filter-demineralizer to remove particulate and dissolved impurities from the reactor coolant. It also removes excess coolant from the reactor system under controlled conditions and provides clean water for the reactor head spray nozzle.

1.2.2.7.2 Fuel Pool Cooling and Cleanup System

The Fuel Pool Cleanup (FPC) System maintains acceptable levels of temperature and clarity and minimizes radioactivity levels of the water in the spent fuel pool, reactor well and dryer/separator pit on top of the containment. The FPC System also maintains the temperature and water level in the service pool and equipment pool. The system includes two heat exchangers, each capable of removing the decay heat generated from an average discharge of spent fuel, and two filter/demineralizers, each unit having the capacity to process the system flow or greater to maintain the desired purity level.

1.2.2.7.3 Suppression Pool Cleanup System

The Suppression Pool Cleanup (SPCU) System provides a continuous purification of the suppression pool water. The system removes impurities by filtration, adsorption, and ion exchange processes. The system consists of a recirculation loop with a pump and isolation valves. Suppression pool water is passed through the Fuel Pool Cooling and Cleanup (FPC) System filter/demineralizers for treatment. Treated water may be diverted to refill the reactor well and the upper pool during refueling outage or provide makeup water to the fuel pool and reactor cooling water (RCW) surge tanks following a seismic event.

1.2.2.8 Control Panels

1.2.2.8.1 Main Control Room Panels

The main control room panel arrangement is summarized in Appendix 18C.

1.2.2.8.2 Control Room Backpanels

The control room backpanels are located in an area adjacent to the main control panels and convenient to the control room crew.

1.2.2.8.3 Radioactive Waste Control Panel

The Radioactive Waste Control Panel System provides the operator interface to the consolidated automatic and remote manual controlling of radioactive waste system


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mechanical, electrical, and chemical process components. It consists of one or more control panels, including panel-mounted meters and displays, CRT displays, status indicating lights, mode and display selector switches, actuating mechanical and electrical components, controllers, and control logic elements and signal conditioning devices and processors. It does not include equipment or process sensors, local panels or equipment-mounted actuators or power controllers.

It is expected that most of the panels of this system will be located in the radioactive waste control room; panels performing the above functions which are located in the main control room shall also belong to this system.

1.2.2.8.4 Local Control Panels

The local control panels provide facilities for the installation and operation of electrical equipment and interconnecting wiring which supports no primary man-machine interface during normal plant operations. Included within the scope of the local control panels shall be the physical panel structure and the wiring associated with the components installed within the panels. The local control panels do not include the major electrical components installed within the panels, which are instead defined and provided as part of the interfacing plant systems.

1.2.2.8.5 Instrument Racks

The instrument racks provide facilities for the installation and operation of locally mounted instrumentation. Included within the scope of the instrument racks shall be the physical structure upon which the instrumentation is mounted and the wiring associated with the instrument installations. The instrument racks do not include the locally mounted instrumentation, which is instead defined and provided as part of the interfacing plant systems.

1.2.2.8.6 Multiplexing System

The Multiplexing System provides redundant and distributed control and instrumentation data communications networks to support the monitoring and control of interfacing plant systems. The system includes electrical devices and circuitry (such as multiplexing units, bus controllers, formatters and data buses) that connect sensors, display devices, controllers, and actuators which are part of these plant systems. The Multiplexing System also includes the associated data acquisition and communication software required to support its function of plant-wide data and control distribution.

1.2.2.8.7 Local Control Boxes

Local control boxes are uniquely identified to provide operational control of an individual piece of electrical equipment.


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1.2.2.9 Nuclear Fuel

1.2.2.9.1 Fuel Assembly

The nuclear fuel assembly contains fissionable material which produces thermal power while maintaining structural integrity. The configuration of the fuel bundle consists of fuel rods, spacers, water rods, upper and lower tie plates. The fuel bundle along with the channel and channel fastener, are assembled into a transportable, interchangeable assembly. The outer envelope of the fuel assembly is square with distinguishing features which provide support, identification, orientation and handling capabilities. The fuel design interface is described in Subsection 4.2.2.1.

1.2.2.9.2 Fuel Channel

The fuel channel encloses the fuel bundle and provides:

(1) A barrier between two parallel coolant flow paths, one for flow inside the fuel bundle and the other for flow in the bypass region between channels.

(2) A bearing surface for the control rod.

(3) Rigidity for the fuel bundle.

The channel fastener attaches the channel to the fuel bundle and, along with the channel spacer buttons, provides channel-to-channel spacing with resilient engagement.

1.2.2.10 Radioactive Waste System

1.2.2.10.1 Radwaste System

1.2.2.10.1.1 Liquid Waste Management System

The Liquid Waste Management System collects, monitors, and treats liquid radioactive wastes for return to the primary system whenever practicable. The radwaste processing equipment is located in the Radwaste Building. Processed waste volumes discharged to the environs are expected to be small. Any discharge is such that concentrations and quantities of radioactive material and other contaminants are in accord with applicable local, state, and federal regulations.

All potentially radioactive liquid wastes are collected in sumps or drain tanks at various locations in the plant. These wastes are transferred to collection tanks in the radwaste facility.

Waste processing is done on a batch basis. Each batch is sampled as necessary in the collection tanks to determine concentrations of radioactivity and other contamination. Equipment drains and other low-conductivity wastes are treated by filtration and


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demineralization and are transferred to the condensate storage tank for reuse. Laundry drain wastes and other detergent wastes of low activity are treated by filtration, sampled and released via the liquid discharge pathway and demineralization and may be released from the plant on a batch basis. Protection against inadvertent release of liquid radioactive waste is provided by design redundancy, instrumentation for the detection and alarm of abnormal conditions, automatic isolation, and administrative controls.

Equipment is selected, arranged, and shielded to permit operation, inspection, and maintenance with minimum radiation exposure to personnel.

1.2.2.10.1.2 Gaseous Waste Management System

The objective of the Gaseous Waste Management System is to process and control the release of gaseous radioactive effluents to the site environs so as to maintain the exposure of persons in unrestricted areas to radioactive gaseous effluents as low as reasonably achievable (10CFR50, Appendix I). This shall be accomplished while maintaining occupational exposure as low as reasonably achievable and without limiting plant operation or availability.

The offgas system provides for holdup and decay of radioactive gases in the offgas from the air ejector system of a nuclear reactor and consists of process equipment along with monitoring instrumentation and control equipment (Section 11.3).

1.2.2.10.1.3 Solid Waste Management System

The Solid Waste Management System provides for the safe handling, packaging, and short-term storage of radioactive solid and concentrated liquid wastes that are produced. Wet waste processed by this system is transferred to the solidification system, where it is solidified in containers. Dry active waste is surveyed and disposed of whenever possible via the provisions of 10CFR20.302 (a). The remaining combustible waste is compacted. Incinerator ash is compacted waste and shipped in containers for offsite disposal.

1.2.2.11 Power Cycle Systems

1.2.2.11.1 Turbine Main Steam System

The Main Steam (MS) System delivers steam from the reactor to the turbine generator, the reheaters, and the steam jet air ejectors (SJAE) from warmup to full-load operation. The MS System also provides steam for the steam seal system and the auxiliary steam system when other steam sources are not available.


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1.2.2.11.2 Condensate, Feedwater and Condensate Air Extraction System

The Condensate and Feedwater System provides a dependable supply of high-quality feedwater to the reactor at the required flow, pressure, and temperature. The condensate pumps take the deaerated condensate from the condenser hotwell and deliver it through the SJAE condenser, gland steam condenser, offgas condenser, condensate filters and demineralizer, and through three parallel strings of four low pressure feedwater heaters to the reactor feed pumps' suction. The reactor feed pumps discharge through two stages of two parallel high pressure feedwater heaters to the reactor. The drains from the high pressure heaters are pumped backward to the suction of the reactors feed pumps.

1.2.2.11.2.1 Main Condenser Evacuation System

The Main Condenser Evacuation System removes the noncondensible gases from the main condenser and discharges them to the offgas system. This system consists of two 100% capacity, multiple-element, multi-stage steam jet air ejectors (SJAE) with intercondensers, for normal station operation, and mechanical vacuum pumps for use during startup.

1.2.2.11.3 Heater, Drain and Vent System

The Heater, Drain and Vent System permits efficient and dependable operation of the heat cycle balance-of-plant equipment and, particularly, the condensate and feedwater regenerative heaters. All process equipment drains and vents are collected and routed to the appropriate points in the cycle and flows are controlled for equipment protection.

1.2.2.11.4 Condensate Purification System

Each unit is served by a 100% capacity condensate cleanup system, consisting of high efficiency filters followed by deep-bed demineralizer vessels designed for parallel operation. One demineralizer vessel is a spare. The condensate cleanup system with instrumentation and automatic controls is designed to ensure a constant supply of high-quality water to the reactor.

1.2.2.11.5 Condensate Filter Facility

The condensate filter facility continuously removes suspended solids by processing the full-flow condensate through high efficiency filters. A fast acting full-flow bypass valve opens on high pressure differential across the filter to protect against sudden loss of condensate flow.


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1.2.2.11.6 Condensate Demineralizer

The condensate demineralizers continuously process condensate to remove dissolved solids to reactor feedwater quality through demineralizers and an additional unit in manual standby. An emergency bypass line protects the equipment, and a demineralizer resin handling and cleaning system is included.

1.2.2.11.7 Main Turbine

The main turbine is a 188.5 rad/s, tandem compound six-flow, reheat steam turbine with 132.08 cm last-stage blades. The turbine generator is equipped with an electro-hydraulic control system and supervisory instruments to monitor performance. The gross electrical output of the turbine generator is approximately 1400 MW.

1.2.2.11.8 Turbine Control System

The Turbine Control System is summarized in Subsection 10.2.2.3.

1.2.2.11.9 Turbine Gland Steam System

The Turbine Gland Steam System provides steam to the turbine shaft glands and the turbine valve stems. The system prevents leakage of air into or radioactive steam out of the turbine shaft and turbine valves. The gland steam condenser collects air and steam mixture, condenses the steam, and discharges the air leakage to the atmosphere via the main vent by a motor-driven blower.

1.2.2.11.10 Turbine Lubricating Oil System

The Turbine Lubricating Oil System supplies oil to turbine-generator bearing lubrication lines and mainly consists of lube oil tank, oil pumps, oil coolers, and oil purifier equipment.

1.2.2.11.11 Moisture Separator Reheater

The moisture separator reheater is summarized in Subsection 10.2.2.2 (Subtopic Moisture Separator Reheater).

1.2.2.11.12 Extraction System

Extraction steam from the high pressure turbine supplies the last stage of feedwater heating and extraction steam from the low pressure turbines supplies the first four stages. An additional low pressure extraction drained directly to the condenser protects the last-stage bucket from erosion induced by water droplets.

1.2.2.11.13 Turbine Bypass System

The turbine bypass system is summarized in Subsection 10.4.4.2.1


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1.2.2.11.14 Reactor Feedwater Pump Driver

Each reactor feedwater pump is driven by an electrical motor driven adjustable speed drive.

1.2.2.11.15 Turbine Auxiliary Steam System

The Turbine Auxiliary Steam System is used when required to supply steam to the steam jet air injectors for condenser deaeration and to the Turbine Gland Seal System, which prevents radioactive steam leakage out of the turbine casings and atmospheric air leakage into the casing at specific operating conditions.

The house boiler steam is a backup to the reactor generated steam during operation and would be used only when reactor steam is unavailable or too radioactive.

1.2.2.11.16 Generator

The generator is a direct-driven, three-phase, 60 Hz, 27 kV, 188.5 rad/s, conductor cooled, synchronous generator rated at approximately 1600 MVA, at 0.90 power factor, 537.4 kPaG hydrogen pressure, and 0.60 short circuit ratio.

1.2.2.11.17 Hydrogen Gas Cooling System

The Hydrogen Gas Cooling System is summarized in Subsection 10.2.2.2 (Subtopic Bulk Hydrogen System).

1.2.2.11.18 Generator Cooling System

The Generator Cooling System includes the hydrogen cooled rotor portion of the Hydrogen Gas Cooling System and the water cooled stator portion of the Turbine Building Cooling Water System.

1.2.2.11.19 Generator Sealing Oil System

The Generator Sealing Oil System prevents hydrogen gas from leaking from the generator. The sealing oil is vacuum-treated to maintain the hydrogen gas purity.

1.2.2.11.20 Exciter

The generator exciter is driven by the main turbine and will have a response ratio that meets the plant voltage regulation requirements and the site specific grid requirements.

Excitation power is provided by the output of a dedicated winding located in the main generator. This output is rectified by the stationary silicon-diode rectifiers. The DC output of the rectifier banks then is applied to the main generator field through the generator collectors.


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1.2.2.11.21 Main Condenser

The main condenser is a multipressure three-shell deaerating type condenser or single pressure design as dictated by the site specific circulating water system and power generating heat sink. During plant operation, steam expanding through the low pressure turbines is directed downward into the main condenser and is condensed. The main condenser also serves as a heat sink for the turbine bypass system, emergency and high level feedwater heater and drain tank dumps, and various other startup drains and relief valve discharges.

1.2.2.11.22 Offgas System

The Offgas System is summarized in Subsection 11.3.

1.2.2.11.23 Circulating Water System

The Circulating Water System provides a continuous supply of cooling water to the condenser to remove the heat rejected by the steam cycle and transfers it to the power cycle heat sink.

1.2.2.11.24 Condenser Cleanup Facility

The condenser cleanup facility removes slime and sludge to prevent vacuum decline of the condenser and to suppress corrosion on the inner surface of the condenser tubes.

1.2.2.12 Station Auxiliary Systems

1.2.2.12.1 Makeup Water System (Preparation)

The Makeup Water System (preparation) is summarized in Subsection 9.2.8.3.

1.2.2.12.1.1 Makeup Water System (Purified)

The Makeup Water System (purified) is summarized in Subsection 9.2.10.2.

1.2.2.12.2 Makeup Water System (Condensate)

The Makeup Water System maintains the required capacity and flow of the condensate for the RCIC and HPCF Systems and maintains the required level in the condenser hotwell. The system also (1) stores and transfers water during refueling and cask storage pool water during fuel shipping cask loading, (2) receives and stores the process effluent from the liquid radwaste system, (3) provides makeup to other plant systems where required, and (4) provides condensate to the Control Rod Drive (CRD) Hydraulic System.

The system consists of a condensate storage tank, three condensate transfer pumps, and the necessary controls and instrumentation.


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1.2.2.12.2.1 Condensate Storage Facilities and Distribution System

The condensate storage tank receives demineralized water from the purified water makeup system and may also receive low conductivity water from the condensate return of the primary loop, from the radwaste disposal system and the condensate system in the Turbine Building.

1.2.2.12.3 Reactor Building Cooling Water System

The Reactor Building Cooling Water (RCW) System provides cooling water to certain designated equipment located in the Reactor Building. Capacity and redundancy is provided in heat exchangers and pumps to ensure adequate performance of the cooling system under all postulated conditions. During loss of offsite power, emergency power for the system is available from the onsite emergency diesel generators. The closed loop design provides a barrier between radioactive systems and the reactor service water discharged to the environment. Heat is removed from the closed loop by the Reactor Service Water System. Radiation monitors are provided to detect contaminated leakage into the closed systems.

1.2.2.12.4 Turbine Building Cooling Water System

The Turbine Building Cooling Water System is summarized in Subsection 9.2.14.2.1.

1.2.2.12.5 HVAC Normal Cooling Water System

The HVAC Normal Cooling Water System provides chilled water to the air supply cooling coils of the reactor building, to the heating/cooling coils in the drywell, and the control building electrical equipment room.

1.2.2.12.6 HVAC Emergency Cooling Water System

The HVAC emergency cooling water system provides chilled water to the cooling coils in the control building essential electrical equipment room, the main control room and the diesel generator electrical equipment areas. The safety-related chilled-water system is designed to meet the requirements of Criterion 19 of 10CFR50.

1.2.2.12.7 Oxygen Injection System

The Oxygen Injection System is summarized in Subsection 9.3.10.2.

1.2.2.12.8 Ultimate Heat Sink

The Ultimate Heat Sink System is summarized in Subsection 9.2.5.3.

1.2.2.12.9 Reactor Service Water System

The Reactor Service Water System is summarized in Subsection 9.2.15.1.3 and 9.2.15.2.3.


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Design Control Document/Tier 2

ABWR

1.2.2.12.10 Turbine Service Water System

The Turbine Service Water System is summarized ion Subsection 9.2.16.1.3 and 9.2.16.2.3.

1.2.2.12.11 Station Service Air System

The Station Service Air System provides a continuous supply of compressed air of suitable quality and pressure for general plant use. The service air compressor discharges into the air receivers and the air is then distributed throughout the plant.

1.2.2.12.12 Instrument Air System

The Instrument Air System is summarized in Subsection 9.3.6.2.

1.2.2.12.13 High Pressure Nitrogen Gas Supply System

Nitrogen gas is normally supplied by the Atmospheric Control System to meet the requirement of (1) the Main Steam System SRV automatic depressurization and relief function accumulators, (2) the main steam isolation valves, and (3) instruments and pneumatic valves using nitrogen in the Reactor Building. When this supply of pressurized nitrogen is not available, the High Pressure Nitrogen Gas Supply (HPIN) System automatically maintains nitrogen pressure to this equipment. The HPIN System consists of high pressure nitrogen storage bottles with piping, valves, instruments, controls and control panel.

1.2.2.12.14 Heating Steam and Condensate Water Return System

The Heating Steam and Condensate Water Return System supplies heating steam from the House Boiler for general plant use and recovers the condensate return to the boiler feedwater tanks. The system consists of piping, valves, condensate recovery set and associated controls and instrumentation.

1.2.2.12.15 House Boiler System

The House Boiler System consists of the house boilers, reboilers, feedwater components, boiler water treatment and control devices. The House Boiler System supplies turbine gland steam and heating steam, including the concentrating tanks and devices of the high conductivity waste equipment.

1.2.2.12.16 Hot Water Heating System

The Hot Water Heating System is a closed-loop hot water supply to the various heating coils of the HVAC systems. The system includes two heat exchangers, surge and chemical addition tanks and associated equipment, controls and instrumentation.


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1.2.2.12.17 Hydrogen Water Chemistry System

The Hydrogen Water Chemistry System is summarized in Subsection 9.3.9.2.

1.2.2.12.18 Zinc Injection System

The Zinc Injection System is summarized in Subsection 9.3.11.1.

1.2.2.12.19 Breathing Air System

The Breathing Air System includes air compressors, dryers, purifiers and a distribution network. This network makes breathing air available in all plant areas where operations or maintenance must be performed and high radioactivity could occur in the ambient air. Special connections are provided to assure that this air is used only for breathing apparatus.

1.2.2.12.20 Sampling System (Includes PASS)

The Process Sampling System is furnished to provide process information that is required to monitor plant and equipment performance and changes to operating parameters. Representative liquid and gas samples are taken automatically and/or manually during plant operation for laboratory or online analyses.

1.2.2.12.21 Freeze Protection System

The Freeze Protection System provides insulation, steam and electrical heating for all external tanks and piping that may freeze during winter weather.

1.2.2.12.22 Iron Injection System

The Iron Injection System consists of an electrolytic iron ion solution generator and means to inject the iron solution into the feedwater system in controlled amounts.

1.2.2.13 Station Electrical Systems

1.2.2.13.1 Electrical Power Distribution System

The unit Class 1E AC power system supplies power to the unit Class 1E loads. The offsite power sources converge at the system. The system includes diesel generators that serve as standby power sources, independent of any onsite or offsite source. Therefore, the system has multiple sources. Furthermore, the system is divided into three divisions, each with its own independent distribution network, diesel generator, and redundant load group. A fourth division battery for the safety logic and control system bus receives charger power from the Division II source.


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1.2.2.13.2 Unit Auxiliary Transformer

The unit auxiliary AC power system supplies power to unit loads that are non-safety-related and uses the main generator as the normal power source with the reserve auxiliary transformer as a backup source. The unit auxiliary transformer steps down the AC power to the 6900V station bus voltage.

1.2.2.13.3 Isolated Phase Bus

The isolated phase bus duct system provides electrical interconnection from the main generator output terminals to the generator breaker and from the generator breaker to the low voltage terminals of the main transformer, and the high voltage terminals of the unit auxiliary transformers. During the time the main generator is off line, the generator breaker is open and power is fed to the unit auxiliary transformers by backfeeding from the main transformer. During startup, the generator breaker is closed at about 7% power to provide power to the main and the unit auxiliary transformers for normal operation of the plant.

A package cooling unit is supplied with the isolated bus duct system.

1.2.2.13.4 Non-Segregated Phase Bus

The non-segregated phase bus provides the electrical interconnection between the unit auxiliary transformers and their associated 6.9 kV metal-clad switchgear.

1.2.2.13.5 Metal-clad Switchgear

The metal-clad switchgear distributes the 6.9 kV power. Circuit breakers are drawout type, stored energy vacuum breakers. The switchgear interrupting rating shall be determined in accordance with requirements of ANSI C37.10.

1.2.2.13.6 Power Center

The power center is summarized in Subsection 8.3.1.1.2.1.

1.2.2.13.7 Motor Control Center

The motor control center is summarized in Subsection 8.3.1.1.2.2.

1.2.2.13.8 Raceway System

The Raceway System is a plant-wide network comprised of metallic cable trays, metallic conduits and supports. Raceways are classified for carrying medium voltage power cables, low voltage power cables, control cables and low level signal/instrumentation cables. Divisional cables are routed in separate cable raceways for each division.

Fiber optic dataways are not restricted to raceway classifications, but would generally be run with control cables due to their common destinations.


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1.2.2.13.9 Grounding Wire

Station grounding and surge protection are discussed in Section 8A.1.

1.2.2.13.10 Electrical Wiring Penetration

Electrical wiring penetrations are described in Subsection 8.3.3.6.1.2 (7).

1.2.2.13.11 Combustion Turbine Generator

The primary function of the Combustion Turbine Generator (CTG) is to act as a standby onsite non-safety power source to feed Plant Investment Protection (PIP) non-safety loads during Loss of Preferred Power (LOPP) events.

The unit also provides an alternate AC power source in case of a station blackout event, as defined by Appendix B of Regulatory Guide 1.155 (Appendix 1C).

1.2.2.13.12 Direct Current Power Supply

The plant has four independent Class 1E and three non-Class 1E 125 VDC power systems.

1.2.2.13.12.1 Unit Auxiliary DC Power System

The Unit Auxiliary DC Power System supplies power to unit DC loads that are non-safety-related. The system consists of three battery chargers, three batteries, and three distribution panels.

1.2.2.13.12.2 Unit Class 1E DC Power System

The Unit Class 1E DC Power System supplies 125 VDC power to the unit Class 1E loads. Battery chargers are the primary power sources. The system, which includes storage batteries that serve as standby power sources, is divided into four divisions, each with its own independent distribution network, battery, and charger.

1.2.2.13.13 Emergency Diesel Generator System

The Emergency Diesel Generator System is supplied by three diesel generators. Each Class 1E division is supplied by a separate diesel generator. There are no provisions for transferring Class 1E buses between standby AC power supplies or supplying more than one engineered safety feature (ESF) from one diesel generator. This one-to-one relationship ensures that a failure of one diesel generator can affect only one ESF division. The diesel generators are housed in the Reactor Building which is a Seismic Category I structure, to comply with applicable NRC and IEEE design guides and criteria.


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1.2.2.13.14 Vital AC Power Supply

1.2.2.13.14.1 Safety System Logic and Control Power System

Four divisions of the Safety System Logic and Control (SSLC) Power System provide an uninterruptible Class 1E source of 120-VAC single-phase control power. The primary power source for the SSLC Power System is the Class 1E AC power system. On loss of AC power, the appropriate divisional battery immediately assumes load without interruption. When AC power is restored, it resumes the load without interruption.

1.2.2.13.14.2 Uninterruptible Power System

The Uninterruptible Power System (UPS) supplies regulated 120 VAC single-phase power to non-Class 1E instrument and control loads which require an uninterruptible source of power. The power sources for the UPS are similar to those for the SSLC, but are non-Class 1E.

1.2.2.13.14.3 Reactor Protection System Alternate Current Power Supply

The Reactor Protection System alternate current power supply is described in Subsection 8.3.1.1.4.2.1.

1.2.2.13.15 Instrument and Control Power Supply

The instrument and control (I&C) power supply provides 120 VAC single-phase power to I&C loads which do not require an uninterruptible power source.

1.2.2.13.16 Communication System

The communication system is summarized in Subsection 9.5.2.

1.2.2.13.17 Lighting and Servicing Power Supply

The design basis for the lighting facilities is the standard for the Illuminating Engineering Society. Special attention is given to areas where proper lighting is imperative during normal and emergency operations. The system design precludes the use of mercury vapor fixtures in the containment and the fuel handling areas. The normal lighting systems are fed from the unit auxiliary transformers. Emergency power is supplied by engineered safety buses backed-up by diesel generators. Normal operation and regular simulated offsite power loss tests verify system integrity.

1.2.2.14 Power Transmission Systems

1.2.2.14.1 Reserve Auxiliary Transformer

The reserve auxiliary transformer provides the alternate preferred feed for the Class 1E buses M/C, E, F, and G. It also provides an alternate feed to non-Class 1E 6.9 kV buses.


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1.2.2.15 Containment and Environmental Control Systems

1.2.2.15.1 Primary Containment System

The ABWR primary containment system design incorporates the drywell/pressure suppression feature of previous BWR containment designs. In fulfilling its design basis as a fission product barrier, the primary containment is a low leakage structure even at the increased pressures that could follow a main steamline rupture or a fluid system line break.

1.2.2.15.1.1 Primary Containment Vessel

The main features of the primary containment design include:

(1) The drywell, a cylindrical steel-lined reinforced concrete structure surrounding the reactor pressure vessel (RPV).

(2) A suppression pool filled with water, which serves as a heat sink during normal operation and accident conditions.

(3) The air space above the suppression pool.

1.2.2.15.2 Containment Internal Structures

The containment internal structures are summarized in Subsections 3.8.3.1 and 6.2.1.1.2.3.

1.2.2.15.3 Reactor Pressure Vessel Pedestal

The reactor pressure vessel (RPV) pedestal is a prefabricated cylindrical steel structure filled with concrete which supports the RPV and is maintained below design temperature by cooling. The pedestal provides drywell connecting vents which lead to the horizontal vent pipes to the suppression pool.

1.2.2.15.4 Standby Gas Treatment System

The Standby Gas Treatment System (SGTS) minimizes exfiltration of contaminated air from the secondary containment to the environment following an accident or abnormal condition which could result in abnormally high airborne radiation in the Reactor Building. Because the fuel storage area is also in the secondary containment, it also can be exhausted to the SGTS.

All safety-related components of the SGTS are operable during loss of offsite power.


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1.2.2.15.5 PCV Pressure and Leak Testing Facility

The PCV pressure and leak testing facility is a special area just outside the containment. It provides instrumentation for conducting the PCV pressure and integrated leak rate tests.

1.2.2.15.6 Atmospheric Control System

The Atmospheric Control System is designed to establish and maintain an inert atmosphere within the primary containment during all plant operating modes except during plant shutdown for refueling or maintenance.

The Atmospheric Control System is summarized in Subsection 6.2.5.2.1.

1.2.2.15.7 Drywell Cooling System

The Drywell Cooling System is summarized in Subsection 9.4.9.2.

1.2.2.15.8 Flammability Control System

A recombiner system is provided to control the concentration of hydrogen and oxygen produced by metal water reaction and radiolysis following a design basis accident in the primary containment.

1.2.2.15.9 Suppression Pool Temperature Monitoring System

The Suppression Pool Temperature Monitoring (SPTM) System is summarized in Subsection 7.6.1.7.1.

1.2.2.16 Structures and Servicing Systems

1.2.2.16.1 Foundation Work

The analytical design and evaluation methods for the containment and Reactor Building walls, slabs and foundation mat and foundation soil are summarized in Subsection 3.8.1.4.1.1.

1.2.2.16.2 Turbine Pedestal

The description for the turbine pedestal is the same as that for foundation work in Subsection 3.8.1.4.1.1.

1.2.2.16.3 Cranes and Hoists

The cranes and hoists are summarized in Subsection 9.1.


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1.2.2.16.4 Elevator

The controlled elevators service the Reactor Building radiation controlled zones from the basemat to the refueling floor. Two additional clean elevators service all elevations of the clean zone.

1.2.2.16.5 Heating, Ventilating and Air Conditioning

The plant environmental control systems control temperature, pressure, humidity, and airborne contamination to ensure the integrity of plant equipment, provide acceptable working conditions for plant personnel, and limit offsite releases of airborne contaminants.

The following environmental systems are provided:

(1) The Control Room Habitability Area HVAC System, consisting of supply, recirculation/exhaust and makeup air cleanup units to ensure the habitability of the control room under normal and abnormal conditions of plant operation.

(2) The Reactor Building Secondary Containment HVAC System maintains a negative pressure in the secondary containment under normal and abnormal operating conditions, thereby isolating the environs from potential leak sources. This system removes heat generated during normal plant operation, shutdown, and refueling periods.

(3) The Drywell Cooling System to remove heat from the drywell generated during normal plant operations including startup, reactor scrams, hot standby, shutdown, and refueling periods.

(4) The R/B Safety-Related Equipment HVAC System to distribute air so that a negative pressure is maintained in the emergency core cooling equipment rooms, thereby isolating the potential airborne contamination in these rooms.

(5) The C/B Safety-Related Equipment Area HVAC System to pressurize the electrical rooms, thus allowing exfiltration of air to the battery rooms for exhaust to the outside atmosphere.

(6) The Spent Fuel Pool Area HVAC System to maintain the refueling floor at a negative pressure with respect to the outside atmosphere to prevent the potential release of airborne contamination.


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(7) The R/B Satey-Related Diesel Generator HVAC System to provide cooling during operation of the diesel generators. A tempered air supply system controls the thermal environment when the diesel generators are not operating.

(8) Coolers in the steam tunnel and ECCS rooms to remove heat generated during operation of the equipment in these rooms.

1.2.2.16.5.1 Potable and Sanitary Water System

The potable and sanitary water includes conceptual site specific designs of a potable water system, a sanitary water system, a sewage treatment system, and a separate non-radioactive drain system. These systems are summarized in Subsections 9.2.4.1.3, 9.2.4.3.2, and 9.3.3.2.3 respectively.

1.2.2.16.6 Fire Protection System

The Fire Protection System is designed to provide an adequate supply of water or chemicals to points throughout the plant where fire protection is required. Diversified fire-alarm and fire-suppression types are selected to suit the particular areas or hazards being protected. Chemical fire-fighting systems are also provided as additions to or in lieu of the water fire-fighting systems. Appropriate instrumentation and controls are provided for the proper operation of the fire detection, annunciation and fire-fighting systems.

1.2.2.16.7 Floor Leakage Detection System

The drainage system is also used to detect abnormal leakage in safety-related equipment rooms and the fuel transfer area.

1.2.2.16.8 Vacuum Sweep System

A portable, submersible-type, underwater vacuum cleaner is provided to assist in removing crud and miscellaneous particulate matter from the pool floors or reactor vessel. The pump and the filter unit are completely submersible for extended periods. The filter “package” is capable of being remotely changed, and the filters will fit into a standard shipping container for offsite burial.

1.2.2.16.9 Decontamination System

The Decontamination System provides areas, equipment and services to support low radiation level decontamination activities. The services may include electrical power, service air, demineralized water, condensate water, radioactive and nonradioactive drains, HVAC and portable shielding.


Rev. 0


1.2.2.16.10 Reactor Building

The Reactor Building includes the containment, drywell, and major portions of the nuclear steam supply system, steam tunnel, refueling area, diesel generators, essential power, non-essential power, emergency core cooling systems, HVAC and supporting systems. The secondary containment is a reinforced concrete building that forms the secondary containment boundary which surrounds the primary containment above the basemat. It permits monitoring and treating all potential radioactive leakage from the primary containment. Treatment consists of HEPA and activated charcoal filtration.

1.2.2.16.11 Turbine Building

The Turbine Building houses all equipment associated with the main turbine generator. Other auxiliary equipment is also located in this building.

1.2.2.16.12 Control Building

The Control Building includes the control room, the computer facility, the cable tunnels, some of the plant essential switchgear, some of the essential power, reactor building water system and the essential HVAC system.

1.2.2.16.13 Radwaste Building

The Radwaste Building houses all equipment associated with the collection and processing of solid and liquid radioactive waste generated by the plant.

1.2.2.16.14 Service Building

The Service Building houses the personnel facilities and portions of the non-essential HVAC System.

1.2.2.17 Yard Structures and Equipment

1.2.2.17.1 Stack

The plant stack is located on the Reactor Building and rises to an elevation of 76 meters above grade level. The stack is a steel shell construction supported by an external steel tubular frame work. The stack vents the Reactor Building, Turbine Building, Radwaste Building, and a small portion of the Control and Service buildings.

1.2.2.17.2 Oil Storage and Transfer System

The major components of this system are the fuel-oil storage tanks, pumps, and day tanks. Each diesel generator has its own individual supply components. Each storage tank is designed to supply the diesel needs during the post-LOCA period, and each day tank has capacity for 8 hours of diesel generator operation at maximum LOCA load demand. Each fuel oil pump is controlled automatically by day-tank level and feeds its


Rev. 0


day tank from the storage tank. Additional fuel oil pumps supply fuel to each diesel fuel manifold from the day tank.

1.2.2.17.3 Site Security

Site Security is summarized in Subsection 13.6.3.1.


1.2.3.1 Plant Design and Aging Management

The COL applicant shall initiate the life cycle management program early in the design process and shall consider the design life requirements as outlined in Subsection 1.2.1.3. In addition, the aging management plan shall cover the structures and components, and the plan shall consider the potential causes of corrosion as outlined in Subsection 1.2.1.3.

Rev. 0

Design Control Docum

ent/Tier 2

General Plant D

escription1.2-41

AB

WR

Figure 1.2-1 Site Plan

No. 1 REACTOR CONTAINMENT 2 REACTOR BUILDING 3 CONTROL BUILDING 4 MAIN STEAM/FEEDWATER TUNNEL 5 TURBINE BUILDING 6 SERVICE BUILDING 7 RADWASTE BUILDING 8 HOUSE BOILER 9 CONDENSATE STORAGE TANK10 UNIT AUXILIARY TRANSFORMERS11 NORMAL SWITCHGEAR12 DIESEL OIL STORAGE TANK (3)13 STACK14 EQUIPMENT ENTRY LOCK15 FIRE PROTECTION WATER STORAGE TANK (2)16 FIRE PROTECTION PUMPHOUSE17 BUNKER FUEL TANK18 COMBUSTION TURBINE GENERATOR19 RADWASTE TUNNELS RB, CB, TB20 DG OIL TRANSFER TUNNEL (3)

7

10

59

3 36

12

113

14

12

15

15

16

11

8

4

12

FACILITY

1817

2

0°

180°


Rev. 0


The following figures are located in Chapter 21 :

Figure 1.2-2 Reactor Building, Arrangement Elevation, Section A-A

Figure 1.2-2a Reactor Building, Arrangement Elevation, Section B-B

Figure 1.2-3 Upper Drywell, Arrangement Elevation, Section A-A

Figure 1.2-3a Upper Drywell, Arrangement Elevation, Section B-B

Figure 1.2-3b Lower Drywell, Arrangement Elevation, Section A-A

Figure 1.2-3c Wetwell, Arrangement Elevation, Sections A-A & B-B

Figure 1.2-4 Reactor Building, Arrangement Plan at Elevation –8200 mm

Figure 1.2-5 Reactor Building, Arrangement Plan at Elevation –1700 mm

Figure 1.2-6 Reactor Building, Arrangement Plan at Elevation 4800/8500 mm

Figure 1.2-7 Not Used

Figure 1.2-8 Reactor Building, Arrangement Plan at Elevation 12300 mm




Figure 1.2-12 Reactor Building, Arrangement Plan at Elevation 31700/38200 mm

Figure 1.2-13a Drywell, Arrangement Plan at Elevation 12300 mm

Figure 1.2-13b Drywell, Arrangement Plan at Elevation 15600 mm

Figure 1.2-13c Drywell, Arrangement Plan at Elevation 18100 mm

Figure 1.2-13d Drywell Steel, Arrangement Plan at Elevation 18100 mm

Figure 1.2-13e Lower Drywell, Arrangement Plan at Elevation –6600 to – 1850 mm

Figure 1.2-13f Lower Drywell, Arrangement Plan at Elevation -1850 to 1750 mm

Figure 1.2-13g Lower Drywell, Arrangement Plan at Elevation 1750 to 4800 mm

Figure 1.2-13h Lower Drywell, Arrangement Plan at Elevation 4800 to 6700 mm

Figure 1.2-13i Wetwell, Arrangement Plan at Elevation – 8200 mm


Rev. 0


Figure 1.2-13j Wetwell, Arrangement Plan at Elevation – 1700 mm

Figure 1.2-13k Wetwell, Arrangement Plan at Elevation 4800 mm

Figure 1.2-14 Control and Service Building, Arrangement Elevation,

Section A-A

Figure 1.2-15 Control and Service Building, Arrangement Elevation,

Section B-B

Figure 1.2-16 Control Building, Arrangement Plan at Elevation –8200 mm

Figure 1.2-17 Control and Service Building, Arrangement Elevation –2150 mm

Figure 1.2-18 Control and Service Building, Arrangement Elevation 3500 mm





Figure 1.2-23a Radwaste Building at Elevation—1500 mm

Figure 1.2-23b Radwaste Building at Elevation 4800 mm

Figure 1.2-23c Radwaste Building at Elevation 12300 mm

Figure 1.2-23d Radwaste Building at Elevation 21000 mm

Figure 1.2-23e Radwaste Building, Section A-A

Figure 1.2-23f Not Used

Figure 1.2-23g Not Used

Figure 1.2-24 Turbine Building, General Arrangement at Elevation 5300 mm




Figure 1.2-28 Turbine Building, General Arrangement, Longitudinal

Section A-A


Rev. 0


Figure 1.2-29 Turbine Building, General Arrangement, Section B-B

Figure 1.2-30 Turbine Building, General Arrangement, Section C-C

Figure 1.2-31 Turbine Building, General Arrangement, Section D-D

Comparison Tables 1.3-1

Rev. 0


1.3 Comparison Tables

This section highlights the principal design features of the plant and compares its major features with those of other BWR facilities. The design of this facility is based on proven technology obtained during the development, design, construction, and operation of BWRs of similar types. The data, performance characteristics, and other information presented here represent a current, firm design.

1.3.1 Nuclear Steam Supply System Design Characteristics

Table 1.3-1 summarizes the design and operating characteristics for the nuclear steam supply systems. Parameters are related to power output for a single plant unless otherwise noted.

1.3.2 Engineered Safety Features Design Characteristics

Table 1.3-2 compares the engineered safety features design characteristics.

1.3.3 Containment Design Characteristics

Table 1.3-3 compares the containment design characteristics.

1.3.4 Structural Design Characteristics

Table 1.3-4 compares the structural design characteristics.

1.3.5 Instrumentation and Electrical Systems Design Characteristics

Table 7.1-1 compares the instrumentation and electrical systems design characteristics.

1.3-2 Comparison Tables

Rev. 0


Table 1.3-1 Comparison of Nuclear Steam Supply System

Design Characteristics

Design1

This Plant

ABWR

278-8722

GESSAR

BWR/6

238-748

NMP-2

BWR/5

251-764

Grand Gulf

BWR/6

251-800

Thermal and Hydraulic (Section 4.4)

Rated power (MWt) 3926 3579 3323 3833

Design power (MWt) (ECCS design basis)

4005 3729 3463 4025

Steam flow rate, Mlb/hr at 420

°F(FW Temp)

16.843 15.40 14.263 16.491

Core coolant flow rate (Mlb/hr) 115.1 104.0 108.5 112.5

Feedwater flow rate (Mlb/hr) 16.807 15.372 14.564 16.455

System pressure, nominal in steam dome (psia)

1040 1040 1020 1040

Average power density (kW/l) 50.6 54.1 49.15 54.1

Maximum linear heat generation rate (kW/ft)

13.4 13.4 13.4 13.4

Average linear heat generation rate (kW/ft)

5.97 5.9 5.40 5.93

Maximum heat flux (Btu/hr/ft2) 361,600 361,600 354,255 361,600

Average Heat flux (Btu/hr/ft2) 161,100 159,500 144,032 160,300

Maximum UO2 temperature (

°F) 3365 3435 3325 3435

Average volumetric fuel temperature (

°F) 2150 2185 2130 2185

Average cladding surface temperature (

°F)566 565 566 565

Minimum critical power ratio (MCPR) 1.17 1.20 1.24 1.20

Coolant enthalpy at core inlet (Btu/lb) 527.7 527.6 527.5 527.9

Core maximum voids within assemblies 75 79 76.2 76

Core average exit quality (% steam) 14.5 14.7 13.1 14.6

Feedwater temperature (

°F) 420 420 420 420

Design power peaking factor

Maximum relative assemble power 1.40 1.40 1.40 1.40

Local peaking factor 1.25 1.13 1.24 1.13

Axial peaking factor 1.40 1.40 1.40 1.40

Total peaking factor 2.43 2.26 2.43 2.26


Rev. 0


Nuclear (first core) (Section 4.3)

Water/UO2 volume ratio (cold) 2.95 2.70 2.55 2.70

Reactivity with strongest control rod out (keff)

<0.99 <0.99 <0.99 <0.99

Dynamic void coefficient (c/%)at core average voids(%) (EOC-rated output)

–5.20c@102%rated output

39.2

–7.16

40.95

–8.57

40.54

–7.14

41.31

Fuel temperature doppler coefficient (c/

°C) (EOC-rated output)–0.360 –0.412 –0.419 –0.396

Initial average U-235 enrichment (%) 2.22 1.90 1.90 1.70

Initial cycle exposure (MWd/short ton) 9950 9138 9200 7500

Fuel Assembly (Section 4.2)

Number of fuel assemblies 872 748 764 800

Fuel rod array 8 x 8 8 x 8 8 x 8 8 x 8

Overall length (inches) 176 176 176 176

Weight of UO2 per assembly (lb)(pellet type)

435 456 466 458

Weight of fuel assembly (lb)(includes channel)

675 697 698 697

Fuel Rods (Section 4.2)

Number of fuel rods per assembly 62 62 63 62

Outside diameter (in.) 0.483 0.483 0.493 0.483

Cladding thickness (in.) 3 0.032 0.032 0.032

Diametral gap, pellet-to-cladding (in.) 3 0.009 0.009 0.009

Length of gas plenum (in.) 3 9.48 14 9.48

Cladding material4 Zircaloy-2 Zircaloy-2 Zircaloy-2 Zircaloy-2

Fuel Pellets (Section 4.2)

Material UO2 UO2 UO2 UO2

Density (% of theoretical) 3 95 95 95

Diameter (in.) 3 0.410 0.416 0.410

Length (in.) 3 0.410 0.420 0.410


Design Characteristics (Continued)

Design1

This Plant

ABWR

278-8722

GESSAR

BWR/6

238-748

NMP-2

BWR/5

251-764

Grand Gulf

BWR/6

251-800


Rev. 0


Fuel Channel (Section 4.2)

Thickness (in.) 0.100 0.120 0.100 0.120

Cross section dimensions (in.) 5.48 x 5.48 5.45 x 5.45 5.48 x 5.48 5.45 x 5.45

Material Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4

Core Assembly (Section 4.2)

Fuel weight as UO2 (lb) 379,221 341,640 265,551 365,693

Core diameter (equivalent) (in.) 203.3 185.2 160.2 191.5

Core height (active fuel) (in.) 146 150 146 150

Reactor Control System (Chapters 4 and 7)

Method of variation of reactor power Movable controlrods and variableforcedcoolant flow

Movablecontrolrods and variableforcedcoolant flow



Number of movable control rods 205 177 185 193

Shape of movable control rods Cruciform Cruciform Cruciform Cruciform

Pitch of movable control rods 12.2 12.0 12.0 12.0

Control material in movable rods B4Cgranulescompactedin SS tubes

B4Cgranulescompactedin SS tubes



Type of control rod drives Bottom entryelectrichydraulicfine motion

Bottomentrylockingpiston



Type of temporary Reactivity control for initial core

Burnablepoison;gadolinia-urania fuel rods






Design1

This Plant

ABWR

278-8722

GESSAR

BWR/6

238-748

NMP-2

BWR/5

251-764

Grand Gulf

BWR/6

251-800


Rev. 0


Incore Neutron Instrumentation (Chapters 4 and 7)

Total number of LPRM detectors 208 164 172 176

Number of incore LPRM penetrations 52 41 43 44

Number of LPRM detectors per penetration

4 4 4 4

Number of SRM penetrations 5 4 4 6

Number of IRM penetrations 105 8 8 8

Total nuclear instrument penetrations 62 53 43 58

Source range monitor range N/A 5 6 6

Intermediate range monitor range N/A 6 6 6

Startup range neutron monitor 8 N/A N/A N/A

Power range monitors range Approximately 1% power to 125% power

Local power range monitors 208 164 172 176

Average power range monitors 4 4 6 8

Number and type of incore neutron source

5 Sb-Be 7 Sb-Be 7 Sb-Be 7 Sb-Be

Reactor Vessel (Section 5.3)

Material Low-alloy steel/stainlessand Ni-Cr-Fe alloy clad

Low-alloysteel/stainlessclad



Design pressure (psig) 1250 1250 1250 1250

Design temperature (

°F) 575 575 575 575

Inside diameter (ft-in.) 23-2 19-10 20-11 20-11

Inside height (ft-in.) 68-11 70-4 72-5 72-7

Minimum base metal thickness(cylindrical section) (in.)

7.50 6.0 6.19 6.19

Minimum cladding thickness (in.) 1/8 1/8 1/8 1/8



Design1

This Plant

ABWR

278-8722

GESSAR

BWR/6

238-748

NMP-2

BWR/5

251-764

Grand Gulf

BWR/6

251-800


Rev. 4


1 English units are utilized in this table since the data obtained from the comparative BWR operating facilities are in English units.

2 Parameters for the core loading in Figure 4.3-1 used in the sensitivity analysis.

3 Proprietary information not included in DCD, (Refer to SSAR Section 1.3, Amendment 32).

4 Free-standing loaded tubes.

5 Shutdown through criticality.

6 Prior criticality to low power.

7 ABWR design utilizes reactor internal pumps (RIPs).

8 Discharge piping from discharge block valve to vessel.

9 Pump and discharge piping to and including discharge block valve.

Reactor Coolant Recirculation (Chapter 5)

Number of recirculation loops 0 2 2 2

Design pressure

inlet leg (psig) N/A7 1250 1650 1250

outlet leg (psig) N/A7 16509

1550816509

155016508

15509

Design temperature ( F) N/A7 575 575 575

Pipe diameter (in.) N/A7 22/24 24 24

Pipe material (ANSI) N/A7 304/316 316k 304/316

Recirculation pump flow rate (gpm) 30,430/pump

42,000 47,200 44,600

Number of jet pumps in reactor N/A7 20 20 24

Main Steamlines (Subsection 5.4.9)

Number of steamlines 4 4 4 4

Design Pressure(psig) 1250 1250 1250 1250

Design temperature ( F) 575 575 575 575

Pipe diameter (in.) 28 26 26/28 28

Pipe material Carbon steel

Carbonsteel

Carbonsteel

Carbonsteel



Design1

This Plant

ABWR

278-8722

GESSAR

BWR/6

238-748

NMP-2

BWR/5

251-764

Grand Gulf

BWR/6

251-800


Rev. 0


Table 1.3-2 Comparison of Engineered Safety Features

Design Characteristics

System/Component1

This Plant

ABWR

278-872

GESSAR

BWR/6

238-748

NMP-2

BWR/5

251-764

Grand Gulf

BWR/6

251-800

Emergency Core Cooling Systems (sized on design power-Section 6.3)

Low Pressure Core Spray Systems2

Number of loops N/A 1 1 1

Flow rate(gpm) N/AN/A

6000 at 122 psid

6350 at128 psid

7000 at122 psid

High Pressure Core Spray System3

Number of loops 2 1 1 1

Flow rate (gpm) 800 at 1177 psid

3200 at100 psid

1550 at 1147 psid

6110 at200 psid

1550 at 1130 psid

6350 at200 psid

1650 at 1147 psid

7000 at200 psid

Reactor Core Isolation Cooling System (Subsection 5.4.6)

Flow rate (gpm) 800 at 165-1192 psia reactorpressure

700 at 165-1192 psia reactorpressure

600 at 1173 psiareactorpressure

800 at 165-1192 psia reactorpressure

Automatic Depressurization System

Number of relief valves 8 8 7 8

Low Pressure Coolant Injection4


Number of pumps 3 3 3 3

Flow rate (gpm/pump) 4200 at40 psid

7100 at20 psid

7450 at 26 psid

7450 at 20 psid


Rev. 0



2 ABWR design utilizes the low pressure flooder mode of the RHR System.

3 ABWR design is a flooder system not a spray system.

4 ABWR design referred to as Low Pressure Flooder.

5 The design of the pumps is, in part, based on the required capacity during the reactor flooding mode.

6 Heat exchanger duty at 20 hours after reactor shutdown.

Auxiliary Systems Residual Heat Removal System (Subsection 5.4.7)

Reactor shutdown cooling mode


Number of pumps 5 3 2 2 2

Flow rate (gpm/pump) 4200 7100 7450 7450

Duty (MBtu/ hr heat exchanger)6 29.0 46.9 41.6 50.0

Number of heat exchangers 3 2 2 2

Primary containment cooling mode flow rate (gpm)

4200 7100 7450 7450

Flow rate (gpm/heat exchanger) 8000 7 7400 25,300 total

Number of pumps 3 loops RCW

7 6 2 at 12,000 gpm 1 at 1300 gpm

Fuel Pool Cooling and Cleanup System (Subsection 9.1.3)

Capacity (MBtu/hr) 6.55 8.0 15.0 11.8

Table 1.3-2 Comparison of Engineered Safety Features


System/Component1

This Plant

ABWR

278-872

GESSAR

BWR/6

238-748

NMP-2

BWR/5

251-764

Grand Gulf

BWR/6

251-800


Rev. 0


Table 1.3-3 Comparison of Containment Design Characteristics

Containment1, 2This Plant

ABWR 278-872

GESSAR

BWR/6 238-748

NMP-2

BWR/5 251-764

Grand Gulf

BWR/6 251-800

Primary

Type Over- and underpressuresuppression

Mark III freestandingsteel with reinforcedconcreteshield building

Over- and under-pressureSuppressionMark II

Mark III reinforcedconcretecontainmentwith steel liner

Construction Reinforced concrete with steel liner; steel structure

Cylindricalfreestandingsteel with ellipsoidalhead

Reinforcedconcrete with steel liner

Reinforcedconcretecylinder with hemisphericalhead; steel lined

Drywell Concrete cylinder

Concretecylinder3

Frustum of cone upper portion

Concretecylinder3

Pressure suppression chamber

Concretecylinder

Freestandingsteel annulus with concrete backing

Cylindricallower portion

Steel lined concreteannulus

Containment internal design pressure (psig)

45 15 45 15

Drywell internal design pressure (psig)

45 30 45 30

Drywell free volume (ft3) 259,563 275,000 303,418 270,000

Pressure suppression chamber free volume (ft3)(HWL)

210,475 1,140,000 192,028 1,400,000

Pressure suppression pool water volume (ft3)(LWL)

126,426 129,600 (upper pool dump = 34,200)

154,794 136,000 (upper pool dump = 72,800)

Submergence of vent pipe below pressure pool surface (ft) (HWL)

11.8 to 20.8 7.5 11.0 max. 7.5 min.

Design temperature of drywell (°F)

340 330 340 330

Downcomer vent pressure loss factor

2.5–3.5 2.5–3.5 1.37 2.5–3.5

Break area/ total vent area 0.01 0.012 0.0108 0.008


Rev. 0



2 Where applicable, containment parameters are based on design rated power.

3 Not part of containment boundary.

4 Not specified.

Primary (Continued)

Calculated maximum drywell pressure after blowdown (psig).

39 23.0 39.7 22.0

Pressure suppression chamber (psig)

26 8.7 34.0 9.0

Initial pressure suppression pool temperature rise (°F)during LOCA

50 50 50 30

Leakage rate (% free volume/day)

0.5 1.0 1.1 0.35

Secondary

Type Controlled leakage

Controlledleakage

Controlledleakageelevatedrelease

Controlledleakage

Construction

Lower levels Reinforced concrete

4 Reinforcedconcrete

Reinforcedconcrete

Upper levels Reinforced concrete

4 Steelsuperstructureand siding

Steelsuperstructureand siding

Roof Reinforced concrete

4 Steel decking Steel decking

Internal design pressure (psig)

0.25 0.25 0.25 0.25

Design in leakage rate (% free volume/day at 0.25 in. H2O)

50 100 100 100

Table 1.3-3 Comparison of Containment Design Characteristics (Continued)

Containment1, 2This Plant

ABWR 278-872

GESSAR

BWR/6 238-748

NMP-2

BWR/5 251-764

Grand Gulf

BWR/6 251-800


Rev. 0



Table 1.3-4 Comparison of Structural Design Characteristics

This Plant

ABWR 278-872

GESSAR

BWR/6 238-748

NMP-2

BWR/5 251-764

Grand Gulf

BWR/6 251-800

Seismic Design (Section 3.7)1

Operating Basis Earthquake

horizontal g None 0.15 0.075 0.075

vertical g None 0.10 0.075 0.05

Safe Shutdown Earthquake

horizontal g 0.3 0.30 0.15 0.15

vertical g 0.3 0.20 0.15 0.10

Wind Design (Subsection 3.3.2)

Translation (mph) 60 70 max.5 min.

70 60

Tangential (mph) 240 290 290 300

/12

Identification of Agents and Contractors 1.4-1

Rev. 0


1.4 Identification of Agents and Contractors

GE has engaged in the development, design, construction, and operation of boiling water reactors since 1955. Table 1.4-1 lists the GE reactors completed, under construction, or on order. As can be seen, GE has substantial experience, knowledge, and capability to design, manufacture, and furnish technical assistance for the installation and startup of reactors.

1.4-2 Identification of Agents and Contractors

Rev. 0


Table 1.4-1 Commercial Nuclear Reactors Completed, Under Construction,

or in Design by General Electric

Station Utility

Rating

(MWe)

Year of

Order

Year of Low

Power

License

Dresden 1 Commonwealth Edison 207 1955 1959

Humboldt Bay Pacific G&E 70 1958 1962

KHAL Germany 15 1958 1961

Garigliano Italy 150 1959 1964

Big Rock Point Consumers Power 72 1959 1963

JPDR Japan 11 1960 1963

KRB Germany 237 1962 1967

Tarapur 1 India 190 1962 1967

Tarapur 2 India 190 1962 1969

GKN Holland 52 1963 1968

Oyster Creek JCP&L 640 1963 1969

Nine Mile Point Niagara Mohawk 610 1963 1969


Pilgrim Boston Edison 670 1965 1972

Millstone 1 NUSCO 652 1965 1970

Tsuruga Japan 340 1965 1970

Nuclenor Spain 440 1965 1971

Fukushima 1 Japan 439 1966 1971

BKW KKM Switzerland 306 1966 1972


Monticello Northern States 548 1966 1970

Quad Cities 1 Commonwealth Edison 789 1966 1972

Browns Ferry 1 TVA 1067 1966 1973


Quad Cities 2 Commonwealth Edison 789 1966 1972

Vermont Yankee Vermont Yankee 515 1966 1972

Peach Bottom 2 Philadelphia Electric 1065 1966 1973

Peach Bottom 3 Philadelphia Electric 1065 1966 1974

FitzPatrick PASNY 821 1968 1974

Shoreham LILCO 820 1967 1984

Identification of Agents and Contractors 1.4-3

Rev. 0


Cooper Nebraska PPD 778 1967 1974


Limerick 1 Philadelphia Electric 1100 1967 1984

Hatch 1 Georgia Power 786 1967 1974


Brunswick 1 Carolina P&L 821 1968 1977

Brunswick 2 Carolina P&L 821 1968 1974

Duane Arnold Iowa ELP 545 1968 1974

Fermi 2 Detroit Edison 1093 1968 1987

Hope Creek 1 PSE&G 1067 1969 1984

Hope Creek 2 PSE&G 1067 1969 1986

Chinshan 1 Taiwan 610 1969 1978

Caorso Italy 822 1969 1977

Hatch 2 Georgia Power 786 1970 1978

La Salle 1 Commonwealth Edison 1078 1970 1982

La Salle 2 Commonwealth Edison 1078 1970 1983

Susquehanna 1 Pennsylvania P&L 1050 1967 1982

Susquehanna 2 Pennsylvania P&L 1050 1968 1984

Chinshan 2 Taiwan 610 1970 1979

Hanford 2 WPPSS 1100 1971 1983

Nine Mile Point 2 Niagara Mohawk 1100 1971 1987

Grand Gulf 1 Mississippi P&L 1250 1971 1982


Tokai Japan 1135 1971 1977

River Bend 1 Gulf States 940 1972 1985

Perry 1 Cleveland Electric 1205 1972 1981

Laguna Verde1 Mexico 660 1972 1988

Leibstadt Switzerland 940 1972 1984

Kuosheng 1 Taiwan 992 1972 1981

Kuosheng 2 Taiwan 992 1972 1982


or in Design by General Electric (Continued)

Station Utility

Rating

(MWe)

Year of

Order

Year of Low

Power

License

1.4-4 Identification of Agents and Contractors

Rev. 0


Clinton 1 Illinois Power 950 1973 1986

Cofrentes Spain 975 1973 1985

Laguna Verde 2 Mexico 660 1973 1994

Kashiwazaki - Kariwa 6

Japan 1300 1987 1996

Kashiwazaki -Kariwa 7

Japan 1300 1987 1997


or in Design by General Electric (Continued)

Station Utility

Rating

(MWe)

Year of

Order

Year of Low

Power

License

Requirements for Further Technical Information 1.5-1

Rev. 0


1.5 Requirements for Further Technical Information

In the December 1986 technical description of the Advanced Boiling Water Reactor (ABWR) GE, in Section 3, provided a description of the test and development program associated with the ABWR. Of the efforts described in that report, all have been satisfactorily completed.

/2

GE Topical Reports and Other Documents 1.6-1

Rev. 0


1.6 GE Topical Reports and Other Documents

Table 1.6-1 is a list of all GE topical reports and any other reports or documents which are referenced in Tier 2 and which contain information utilized for the ABWR.

1.6-2 GE Topical Reports and Other Documents

Rev. 0


Table 1.6-1 Referenced Reports

Report No. Title

Tier 2

Section No.

22A7007 “GESSAR II, 238 Nuclear Island, BWR/6 Standard Plant, General Electric Company”, March 1980, & Amendments 1-21.

3.719.219.319D.319D.719E.219E.3

APED-5750 “Design and Performance of General Electric Boiling Water Reactor Main Steam Line Isolation Valves”, General Electric Company, Atomic Power Equipment Department, March 1969.

5.4

NEDO-10029 An Analytic Study on Brittle Fracture of GE-BWR Vessel Subject to the Design Basis Accident, June 1969.

5.3

NEDO-10299A H.T. Kim, “Core Flow Distribution in a Modern BWR as Measured in Monticello”, October 1976.

4.4

NEDO-10527 C.J. Paone and J.A. Woolley, “Rod Drop Accident Analysis for Large Boiling Water Reactors”, Licensing Topical Report, March 1972.

15.4

NEDO-10585 F.G. Brutchscy, et al., “Behavior of Iodine in Reactor Water During Plant Shutdown and Startup”, August 1972.

15.2

NEDO-10722 H.T. Kim, “Core Flow Distribution in a Large Boiling Water Reactor as Measured in Quad Cities Unit 1”, December 1972.

4.4

NEDO-10722A H.T. Kim, “Core Flow Distribution in a Large Boiling Water Reactor as Measured in Quad Cities Unit 1”, August 1976.

4.4

NEDO-10802 - A R.B. Linford, “Analytical Methods of Plant Transients Evaluations for the GE BWR”, December 1986.

4.4

NEDO-10802-01A R.B. Linford, “Analytical Methods of Plant Transients Evaluations for the GE BWR”, Amendment 1, December 1986.

4.4

NEDO-10802-02A R.B. Linford, “Analytical Methods of Plant Transients Evaluations for the GE BWR”, Amendment 2, December 1986.

4.4

NEDO-10871 J.M. Skarpelos and R.S. Gilbert, “Technical Derivation of BWR 1971 Design Basis Radioactive Material Source Terms”, March 1973.

11.1

NEDO-10958-A H.T. Kim, “General Electric Thermal Analysis Basis (GETAB): Data, Correlation and Design Applications (LTR)”, January 1977.

4.44B

NEDO-11209-04-A “GE Nuclear Energy Quality Assurance Program Description”, the latest NRC-accepted version.

17.1


Rev. 0


NEDE13426P “T. R. McIntyre, et. al., Mark III Conformatory Test Program -1/3 Scale Impact Tests - Test Series 5805”, August 1975.

3B

NEDO-20206 D.R. Rogers, “BWR Turbine Equipment N-16 Radiation Shielding Studies”, December 1973.

12.2

NEDO-20340 J. Carew, “Process Computer Performance Evaluation Accuracy” June 1974.

4.3

NEDO-20533 W.J.Bilanin, “The GE Mark III Pressure Suppression Containment Analytical Model”, June 1974.

6.2

NEDO-20533-1 W.J.Bilanin, “The GE Mark III Pressure Suppression Containment Analytical Model”, Supplement 1, September 1975

6.2

NEDE-20566-A “General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix K”, September 1986.

6.3

NEDM-20609-01 P.P. Stancavage and D.G. Abbott, “Liquid Discharge Doses LIDSR Code”, August 1976.

12.2

NEDO-20953A J.A. Woolley, “Three-Dimensional BWR Core Simulator”, January 1977.

4A.4

NEDO-21052 F.J. Moody, “Maximum Discharge Rate of Liquid-Vapor Mixtures from Vessels”, General Electric Company, September 1975.

6.2

NEDO-21143-1 H. Careway, V. Nguyen, and P. Stancavege, “RadiologicalAccident-The CONACO3 CODE”, December 1981.

15.215.6

NEDO-21159 “Airborne Releases from BWRs for Environmental Impact Evaluations”, March 1976.

11.1

NEDO-21159 “Airborne Releases from BWRs for Environmental Impact Evaluations” - Amendment 2 - Iodine

12.2

NEDE-21175-P “BWR/6 Fuel Assembly Evaluation of Combined Safe Shutdown Earthquake (SSE) and Loss-of-Coolant Accident (LOCA) Loadings”, November 1976.

3.9

NEDC-21215 “Brunswick Steam Electric Plant Unit 1 Safety Analysis Report for Plant Modifications to Eliminate Significant In-Core Vibrations”, March 1976.

4.4

NEDC-21251 J. Charnley, “KKM Safety Analysis Report”, April 1976. 4.4

NEDE-21354-P “BWR Fuel Channel Mechanical Design and Deflection”, September 1976.

3.9

NEDE-21471-1 L. Lasher,et. al, “Analytical Model for Estimating Drag Forces on Rigid Submerged Structures Caused Supplement for X-Quencher Air Discharge”, October 1979.

3B

Table 1.6-1 Referenced Reports (Continued)

Report No. Title

Tier 2

Section No.


Rev. 0


NEDO-21471 F. Moody, “Analytical Model for Estimating Drag Forces on Rigid Submerged Structures Caused by a LOCA”, September 1977.

3B

NEDO-21506 “Stability and Dynamic Forces of the GE Boiling Water Reactor (LTR)”, October 1976.

4.1

NEDE-21514 - 1&2 “BWR Scram System Reliability Analysis”, December 1976, General Electric Company.

19D.6

NEDE-21526 J. Dougherty, “SCAM - Subcompartment Analysis Method”, January 1977.

6.2

NEDE-21544-P R.J Ernst, et. al., “Mark II Pressure Suppression Containment Systems: An Analytical Model of the Pool Swell Phenomenon”, December 1977.

3B

NEDO-21778-A “Transient Pressure Rises Affecting Fracture Toughness Requirements for Boiling Water Reactors”, December 1978.

5.3

NEDO-21985 “Functional Capability Criteria for Essential Mark II Piping”, September 1978, prepared by Battelle Columbus Laboratories for General Electric Company.

3.9

NEDE-22056 “Failure Rate Data Manual for GE BWR Components”, Rev. 2 January 17, 1986, Class III, General Electric Company.

19.319D.319E.2

NEDO-22155 “GE Report, Generation and Mitigation of Combustible Gas Mixtures in Inerted BWR Mark I Containments”, June 1982.

6.2

NEDE-22277-P-I G. A. Watford, “Compliance of the GE BWR Fuel Design to Stability Licensing Criteria”, October 1984.

20.3.7

NEDE-23819 P.D. Knecht, “BWR/6 Drywell and Containment Maintenance and Testing Access Time Estimates”, May 1978.

12.4

NEDE-23996-1 P.D. Knecht, “Maintenance Access Time Estimates, BWR/6 Auxiliary and Fuel Buildings”, May 1979.

12.4

NEDE-23996-2 A. Chappori, “Maintenances Access Time Estimates, BWR/6 Radwaste Building”, May 1979.

12.4

NEDO-24057 “Assessment of Reactor Internals Vibration in BWR/4 and BWR/5 Plants”, November 1977.

3.9

NEDO-24057-P “Assessment of Reactor Internals Vibration in BWR/4 and BWR/5 Plants”, November 1977.

3.9

NEDE-24131 “Basis for 8x8 Retrofit Fuel Thermal Analysis Application”, September 1978.

4D.2

NEDO-24154 “Qualification of the One-Dimensional Core Transient Model for BWRs”, Vol. 1 & 2, October 1978.

4.4


Report No. Title

Tier 2

Section No.


Rev. 0


NEDO-24154-P “Qualification of the One-Dimensional Core Transient Model for BWRs”, Vol. 3 October 1978. (Proprietary)

4.4

NEDE-24222 J. Weiss, “Assessment of BWR Mitigation of ATWS”, December 1979.

15E19.3

NEDE-24302-P “Mark II Containment Program, Generic Chugging Load Definition Report”, April 1981.

3B

NEDE-24326-1-P “General Electric Environmental Qualification Program”, Proprietary Document, January 1983.

3.93.11

NEDE-24351 D. Hale, “Fatigue Crack Growth in Piping and RPV Steels in Simulated BWR Water Environment Update”, July 1981.

3E

NEDE-24679 “Study of Advanced BWR Features, Plant Definition/Feasibility Results”, Vol.III, Part G, October 1979.

12.4

NEDO-24708 P. W. Marriot, “Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors”, August 1979.

7.3

NEDE-25100-P “Mark II Containment Supporting Program, Caorso SRV Discharge Tests Phase I Test Report”, May 1979.

3B

NEDE-25118 “Mark II Containment Supporting Program, Caorso SRV Discharge Tests Phase II ATR”, August 1979.

3B

NEDO-25132A E. W. Bradley, “Gamma & Beta Dose to Man from Noble Gas Release to the Atmosphere GEMAN Code”, April 1980.

12.2

NEDO-25153 L. E. Lasher, “Analytical Model for Estimating Drag Forces on Rigid Structures Caused by Steam Condensation and Chugging”, July 1979.

3B

NEDE-25250 A. Javid, “Generic Criteria for High Frequency Cutoff of BWR Equipment”, January 1980. (Proprietary)

3.9

NEDO-25257 E. W. Bradley and V. D. Nguyen, “Radiation Exposure from Airborne Effluents-the REFAE Code”, July 1980.

12.2

NEDE-25273 F. T. Dodge, “Scaling Study of the General Electric Pressure Suppression Test Facility - Mark III Long Range Program, Task 2.2.1”, SwRI, March 1980. (Proprietary)

3B

NEDE-30090 “Alto Lazio Station Reliability Analysis”, December 1984 19D.6

NEDO-30130-A Bill Zarbis, “Steady-State Nuclear Methods”, May 1985. (Proprietary)

4.34.4

NEDC-30259 H.A. Careway, D.B. Townsend, B.W. Shaffer, “A Technique for Evaluation of BWR MSIV Leakage Contribution to Radiological Dose Rate Calculations”, September 1983.

15.6

NEDE-30637 B.M. Gordon, “Corrosion and Corrosion Control in BWRs”, December 1984.

5.2


Report No. Title

Tier 2

Section No.


Rev. 0


NEDE-30640 “Evaluation of Proposed Modification to the GESSAR II Design”, Class III, June 1984.

19P

NEDO-30832 J.E. Torbeck, “Elimination of Limit on BWR Suppression Pool Temperature for SRV Discharge With Quenchers”, December 1984.

3B

NEDC-30851P-A W. P. Sullivan, “Technical Specification Improvement Analyses for BWR Reactor Protection System”, March 1988.

19D.6

NEDE-31096-A “GE Licensing Topical Report ATWS Response to NRC ATWS Rule 10CFR 50.62”, February 1987.

19B.2

NEDE-31152-P “GE Bundle Designs”, December 1988. 4.2

NEDO-31331 Gerry Burnette, “BWR Owner’s Group Emergency Procedure Guidelines”, March 1987.

18A

NEDC-31336 Julie Leong, “General Electric Instrument Setpoint Methodology”, October 1986.

7.3

NEDC-31393 “ABWR Containment Horizontal Vent Confirmatory Test, Part I”, March 1987.

3B

NEDO-31439 C. VonDamm, “The Nuclear Measurement Analysis & Control Wide Range Neutron Monitoring System (NUMAC-WRNMS)”, May 1987

20.3

NEDC-31858P Louis Lee, “BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control System”, 1991

15.6

NEDE-31906-P A. Chung, “Laguna Verde Unit I Reactor Internals Vibration Measurement”, January 1991.

7.4

NEDO-31960 Glen Watford, “BWR Owners’ Group Long-Term Stability Solutions Licensing Methodology”, June 1991.

4.4

NEDC-32267P “ABWR Project Application Engineering Organization and Procedures Manual”, December 1993.

17.1


Report No. Title

Tier 2

Section No.

Drawings 1.7-1

Rev. 0


1.7 Drawings

1.7.1 Piping and Instrumentation and Process Flow Drawings

Table 1.7-1 contains a list of system piping and instrumentation diagrams (P&ID) and process flow diagrams (PFD) provided in Tier 2. Figure 1.7-1, sheets 1 and 2 define the symbols used on these drawings.

1.7.2 Instrument, Control and Electrical Drawings

Interlock block diagrams (IBD), instrument engineering diagrams (IED) and single-line diagrams (SLD) are listed in Table 1.7-2. Figure 1.7-2 defines the graphic symbols used in the IBDs.

1.7.3 ASME Standard Units to Preferred Metric Conversion Factors

The ASME standard units are applied with the numerical values converted to the preferred metric units system as listed in Table 1.7-3.

1.7.4 Preferred Metric Conversion to ASME Standard Units

Selected flow, pressure, temperature, and length preferred metric units are converted to ASME standard units as listed in Table 1.7-4.

1.7.5 Drawing Standards

Guidelines for identifying systems, facilities, equipment types, and numbers and for drawing P&IDs and PFDs are treated in Table 1.7-5.


1.7.6.1 P&ID Pipe Schedules

COL applicants shall complete P&ID pipe schedules indicated as: COL applicant.

1.7-2 Drawings

Rev. 0


Table 1.7-1 Piping and Instrumentation and Process Flow Diagrams

Tier 2 Fig. No. Title Type

4.6-8 CRD System P&ID

4.6-9 CRD System PFD

5.1-3 Nuclear Boiler System P&ID

5.4-4 Reactor Recirculation System P&ID

5.4-5 Reactor Recirculation System PFD

5.4-8 Reactor Core Isolation Cooling System P&ID

5.4-9 Reactor Core Isolation Cooling System PFD

5.4-10 Residual Heat Removal System P&ID

5.4-11 Residual Heat Removal System PFD

5.4-12 Reactor Water Cleanup System P&ID

5.4-13 Reactor Water Cleanup System PFD

6.2-39 Atmospheric Control System P&ID

6.2-40 Flammability Control System P&ID

6.3-1 High Pressure Core Flooder System PFD

6.3-7 High Pressure Core Flooder System P&ID

6.5-1 Standby Gas Treatment System P&ID

6.7-1 High Pressure Nitrogen Gas Supply System P&ID

9.1-1 Fuel Pool Cooling and Cleanup System P&ID

9.1-2 Fuel Pool Cooling and Cleanup System PFD

9.2-1 Reactor Building Cooling Water System P&ID

9.2-2 HVAC Normal Cooling Water System P&ID

9.2-3 HVAC Emergency Cooling Water System P&ID

9.2-4 Makeup Water System (Condensate) P&ID

9.2-5 Makeup Water System (Purified) P&ID

9.2-7 Reactor Service Water System P&ID

9.3-1 Standby Liquid Control System P&ID

9.3-1A Standby Liquid Control System PFD

9.3-6 Instrument Air System P&ID

9.3-7 Service Air System P&ID

9.4-1 Control Building HVAC PFD

9.4-8 Drywell Cooling System P&ID

9.5-1 Suppression Pool Cleanup System P&ID

Drawings 1.7-3

Rev. 0


11.2-1 Radwaste System PFDSimplified

11.3-1 Offgas System PFD

11.3-2 Offgas System P&ID

Table 1.7-1 Piping and Instrumentation and Process Flow Diagrams (Continued)


1.7-4 Drawings

Rev. 0


Table 1.7-2 Instrument Engineering, Interlock Block

and Single-Line Diagrams


5.2-8 Leak Detection and Isolation System IED

5.4-14 Reactor Water Cleanup System IBD

7.2-9 Reactor Protection System IED

7.2-10 Reactor Protection System IBD

7.3-1 High Pressure Core Flooder System IBD

7.3-2 Nuclear Boiler System IBD

7.3-3 Reactor Core Isolation Cooling System IBD

7.3-4 Residual Heat Removal System IBD

7.3-5 Leak Detection and Isolation System IBD

7.3-6 Standby Gas Treatment System IBD

7.3-7 Reactor Building Cooling Water/Reactor Service Water System

IBD

7.3-9 HVAC Emergency Cooling Water System IBD

7.3-10 High Pressure Nitrogen Gas System IBD

7.4-1 Standby Liquid Control System IBD

7.4-2 Remote Shutdown System IED

7.4-3 Remote Shutdown System IBD

7.6-1 Neutron Monitoring System IED

7.6-2 Neutron Monitoring System IBD

7.6-5 Process Radiation Monitoring System IED

7.6-7 Containment Atmosphere Monitoring System IED

7.6-8 Containment Atmosphere Monitoring System IBD

7.6-11 Suppression Pool Temperature Monitoring System

IED

7.6-12 Suppression Pool Temperature Monitoring System

IBD

7.7-2 Rod Control and Information System IED

7.7-3 Rod Control and Information System IBD

7.7-4 Control Rod Drive System IBD

7.7-5 Recirculation Flow Control System IED

7.7-7 Recirculation Flow Control System IBD

7.7-8 Feedwater Control System IED

Drawings 1.7-5

Rev. 0


7.7-9 Feedwater Control System IBD

7.7-12 Steam Bypass and Pressure Control System IED

7.7-13 Steam Bypass and Pressure Control System IBD

7.7-14 Fuel Pool Cooling and Cleanup System IBD

8.3-1 Electrical Power Distribution System SLD

8.3-2 Instrument and Control Power Supply System SLD

8.3-3 Plant Vital AC Power Supply System SLD

8.3-4 Plant Vital DC Power Supply System SLD

Table 1.7-2 Instrument Engineering, Interlock Block

and Single-Line Diagrams (Continued)


1.7-6 Drawings

Rev. 0


Table 1.7-3 Conversion to ASME Standard Units

From To convert to Divide by

(1) Pressure/Stress

kilopascal 1 Pound/Square Inch 6.894757

kilopascal 1 Atmosphere (STD) 101.325

kilopascal 1 Foot of Water (39.2

°F) 2.98898

kilopascal 1 Inch of Water (60

°F) 0.24884

kilopascal 1 Inch of HG (32

°F) 3.38638

(2) Force/Weight

newton 1 Pound - force 4.448222

kilogram 1 Ton (Short) 907.1847

kilogram 1 Tons (Long) 1016.047

(3) Heat/Energy

joule 1 Btu 1055.056

joule 1 Calorie 4.1868

kilowatt-hour 1 Btu 0.0002930711

kilowatts 1 Horsepower (U.K) 0.7456999

kilowatt-hour 1 Horsepower-Hour 0.7456999

kilowatt 1 Btu/Min 0.0175725

joule/gram 1 Btu/Pound 2.326

(4) Length

millimeter 1 Inch 25.4

centimeter 1 Inch 2.54

meter 1 Inch 0.0254

meter 1 Foot 0.3048

centimeter 1 Foot 30.48

meter 1 Mile 1609.344

kilometer 1 Mile 1.609344

(5) Volume

liter 1 Cubic Inch 0.01638706

cubic centimeter 1 Cubic Inch 16.38706

cubic meter 1 Cubic Foot 0.02831685

Drawings 1.7-7

Rev. 0


From To Convert to Divide by

cubic centimeter 1 Cubic Foot 28316.85

liter 1 Cubic Foot 28.31685

cubic meter 1 Cubic Yard 0.7645549

liter 1 Gallon (US) 3.785412

cubic centimeter 1 Gallon (US) 3785.412

E-03 cubic centimeter 1 Gallon (US) 3.785412

(6) Volume Per Unit Time

cubic centimeter/s 1 Cubic Foot/Min 471.9474

cubic meter/h 1 Cubic Foot/Min 1.69901

liter/s 1 Cubic Foot/Min 0.4719474

cubic meter/s 1 Cubic Foot/Sec 0.02831685

E-05 cubic meter/s 1 Gallon/Min (US) 6.30902

cubic meter/h 1 Gallon/Min (US) 0.22712

liter/s (101.325 kPaA, 15.56

°C)1 STD CFM (14.696 psia, 60oF) 0.4474

cubic meter/h (101.325 kPaA, 15.56

°C)1 STD CFM (14.696 psia, 60oF) 1.608

(7) Velocity

centimeter/s 1 Foot/Sec 30.48

centimeter/s 1 Foot/Min 0.508

meter/s 1 Foot/Min 0.00508

meter/min 1 Foot/Min 0.3048

centimeter/s 1 Inches/Sec 2.45

(8) Area

square centimeter 1 Square Inch 6.4516

E-04 square meter 1 Square Inch 6.4516

square centimeter 1 Square Foot 929.0304

E-02 square meter 1 Square Foot 9.290304

(9) Torque

newton-meter 1 Foot Pound 1.355818

(10) Mass Per Unit Time

kilogram/s 1 Pound/Sec 0.4535924

Table 1.7-3 Conversion to ASME Standard Units (Continued)

1.7-8 Drawings

Rev. 0


Note:

Rounding of Calculated values per Appendix C of ANSI/IEEE Std 268.

From To Convert to Divide by

kilogram/min 1 Pound/Min 0.4535924

kilogram/h 1 Pound/Min 27.215544

(11) Mass Per Unit Volume

kilogram/cubic meter 1 Pound/Cubic Inch 27679.90

kilogram/cubic meter 1 Pound/Cubic Foot 16.01846

kilogram/cubic centimeter 1 Pound/Cubic Inch 0.0276799

liter/s 1 Gallon/Min 0.0630902

(12) Dynamic Viscosity

Pa•s 1 Pound-Sec/Sq Ft 47.88026

(13) Specific Heat/Heat Transfer

joule/kilogram kelvin 1 Btu/Pound-Deg F 4168.8

watt/square meter kelvin 1 Btu/Hr-Sq Ft-Deg F 5.678263

watt/square meter kelvin 1 Btu/Sec-Sq Ft-Deg F 2.044175E+4

watt/square meter 1 Btu/Hr-Sq Ft 3.154591

(14) Temperature

degrees celsius Degrees Fahrenheit T

°F=T

°Cx1.8+32

Degree C Increment 1 Degree F Increment 0.555556

(15) Electricity

coulomb 1 ampere hour 3600

seimens/meter 1 mho/centimeter 100

(16) Light

candels/square meter 1 candela/square inch 1550.003

lux 1 footcandle 10.76391

(17) Radiation

megabequerel 1 curie 37,000

gray 1 rad 0.01

sievert 1 rem 0.01

Table 1.7-3 Conversion to ASME Standard Units (Continued)

Drawings 1.7-9

Rev. 0


Table 1.7-4 Conversion Tables—Metric to ASME Standard Units

Flow-Volume Per Unit Time

m3/h gal/min m3/h gal/min m3/h gal/min m3/h gal/min

1 4.4 10 44 100 440 1000 4402

2 8.8 20 88 200 881 2000 8805

3 13.2 30 132 300 1321 3000 13207

4 17.6 40 176 400 1761 4000 17610

5 22.0 50 220 500 2201 5000 22012

6 26.4 60 264 600 2641 6000 26414

7 30.8 70 308 700 3082 7000 30817

8 35.2 80 352 800 3522 8000 35219

9 39.6 90 396 900 3962 9000 39621

Temperature

°C

°F

°C

°F

°C

°F

°C

°F

0.1 32.18 1 33.8 10 50 100 212

0.2 32.36 2 35.6 20 68 200 392

0.3 32.54 3 37.4 30 86 300 572

0.4 32.72 4 39.2 40 104 400 752

0.5 32.90 5 41.0 50 122 500 932

0.6 33.08 6 42.8 60 140 600 1112

0.7 33.26 7 44.6 70 158 700 1292

0.8 33.44 8 46.4 80 176 800 1472

Pressure

kPa psi kPa psi kPa psi kPa psi

1 0.145 10 1.45 100 14.51 1000 145.1

2 0.290 20 2.90 200 29.01 2000 290.1

3 0.435 30 4.35 300 43.52 3000 435.2

4 0.580 40 5.80 400 58.02 4000 580.2

5 0.725 50 7.25 500 72.53 5000 725.3

6 0.870 60 8.70 600 87.03 6000 870.3

7 1.015 70 10.15 700 101.54 7000 1015.4

8 1.160 80 11.60 800 116.04 8000 1160.4

9 1.306 90 13.06 900 130.55 9000 1305.5

1.7-10 Drawings

Rev. 0


Length

cm inch cm inch m ft m ft

0.01 0.004 0.1 0.039 1 3.28 10 32.81

0.02 0.008 0.2 0.079 2 6.56 20 65.62

0.03 0.012 0.3 0.118 3 9.84 30 98.43

0.04 0.016 0.4 0.157 4 13.12 40 131.2

0.05 0.020 0.5 0.197 5 16.40 50 164.0

0.06 0.024 0.6 0.236 6 19.69 60 196.9

0.07 0.028 0.7 0.276 7 22.97 70 229.7

0.08 0.032 0.8 0.315 8 26.25 80 262.5

0.09 0.035 0.9 0.354 9 29.53 90 295.3

Table 1.7-4 Conversion Tables—Metric to ASME Standard Units (Continued)

Drawings 1.7-11

Rev. 0


Note: The following letters are not used: I, L, M, O, Q, V, X, Z

Table 1.7-5 Drawing Standards

1.0 Equipment Identification

1.1 System Groups

The plant systems and facilities are divided into several major groups. Each group is represented by a single alphabetical letter as follows:

A Plant in generalB Reactor steam-generating systemsC Control systemsD Radiation monitoring systemsE Core cooling systemsF Reactor handling equipmentG Reactor auxiliary systemsH Control panelsJ FuelsK Waste-processing systemsN Plant main systemsP Plant auxiliary systemsR Onsite electrical systemsS Power transmission and receiving systemsT Reactor containment vessel and ancillary facilitiesU Various buildings and ancillary facilitiesW Water intake facilities and ancillary facilitiesY Other facilities on the grounds

Equipment Suffix #2 (Section 1.6)

Equipment Suffix #1 (Section 1.5)

Equipment Number (Section 1.4)

Equipment Type (Section 1.3)

System Number (Section 1.2)

System Group (Section 1.1)

1.7-12 Drawings

Rev. 0


Table 1.7-5 Drawing Standards (Continued)

1.2 System Numbers

The system number for each system or facility consists of a two-digit number. Table 3.2-1 shows the system group and system numbers (MPL numbers) for each system and facility.

1.3 Equipment Type

The equipment type is represented by from one to four alphabetical letters as follows:

Mechanical Equipment

Identifying

Letter Description

A Tanks Such as collection tanks, sample tanks, surge tanks, precoat tanks, backwashing tanks, sludge and resin tanks, other tanks, lining vats

B Heat transferequipment

Various types of heat exchangers, coolers, condensers, heaters

C Rotating equipment Such as various types of pumps and prime movers, fans and blowers, generators, exciters

D Other equipment Such as reactor pressure vessel, reactor internals, steam separators, dryers, control rod drive mechanisms, hydraulic control units, control rods, flow-limiting orifices, strainers, filters, demineralizers, agitators, extractors, ejectors, dispersers, and other types of equipment

E Tools and servicing equipment

F Valves and their operators (where supplied)

G Pipes, hangers and supports

H Insulation

Structural Equipment

Identifying

Letter Description

U Foundation and supporting structure

V Steel structures

W Structural concrete and reinforcement bars

X Equipment structures such as flues, chimneys, ducts, louvers, and cable trays

Drawings 1.7-13

Rev. 0



Electrical Equipment

Identifying

Letter DescriptionIdentifying

Letter

Description

J Electrical equipment-buses, transformers, power supply facilities

MV/I Millivolt/current converters

K Auxiliary relays O/E Optic/electric converters

L Limiters P/E Pneumatic-electric converters (including air-pressure-to-current and air-pressure-to-voltage converters)

P Panels and racks R/I Resistance/current converters

S Operation switch RMC Remote controllers

T Timing relays RMS Remote operating switches

Z Complicated controllers such as ratio setters, function generators, division/multiplication calculators, time-lag calculators, addition/subtraction calculators. All microprocessor based algorithms.

R/P Resistance/pneumatic converters

AM Analog memory RY Relay modules

D/D Converters SQRT Square-root calculators

E/O Electric/optic converters SRU Resistance units

E/P Electropneumatic converters (including current-to-air-pressure and voltage-to-air-pressure convertors)

S/S Selector switch

E/S Power supply for instrumentation SSA Selector-selector switch automatic

E/T Relay terminal boards TDS Time delay switches

I/O I/O module TMC Cycle timers

I/V Current/voltage converters TPR Program timers

M/A Manual and manual/automatic controllers

V/V Voltage/voltage converters

MRY Deviation monitor

1.7-14 Drawings

Rev. 0



1.3.1 Instrumentation

The identification for “equipment type” provides information about the measured variable as well as the instrument function (Figure 1.7-1, sheet 2).

1.4 Equipment Number

Mechanical, structural and electrical equipment is numbered from 001 to 999 within the system for each equipment type.

1.4.1 Flow Direction Numbering Method

Equipment numbers are assigned in the direction of flow starting from the reactor vessel (or the upstream flow boundary) and moving in sequence from upstream to downstream. In systems which have two flow paths, the main flow path takes priority. The “Flow-Direction” method takes priority over the “Alternate” method below.

1.4.2 Alternate Numbering Method

For items having the same “Equipment Type” with different specifications and arranged in parallel, the equipment numbers are assigned according to equipment layout following a priority according to the direction, either from north to south or from the sea to the mountains. The north-to-south direction takes priority over the sea-to-mountain direction. However, within a system, the degree of importance of the individual pieces of equipment takes priority over the aforesaid rule, and the numbers are assigned in order of priority from the more important pieces of equipment. In the case in which the items are in parallel and are arranged above and below each other (e.g., upper and lower floors), the priority in numbering from more important to less important supersedes numbering from upper to lower floor.

1.4.3 Rules for Adding and Eliminating Equipment

When equipment is added to a system as the design progresses, the sequential numbers in the system upstream and downstream of the added equipment are not changed, and the added equipment is given the number following the last of the sequential numbers of the equipment at that time. When equipment is eliminated, its equipment number shall not be used again, and the numbers of the equipment on the downstream side remain unchanged.

Drawings 1.7-15

Rev. 0



1.4.4 Valve Numbering

Valves are divided into three categories—(1) Process Valves, (2) Drain Valves and Vent Valves, and (3) Instrument Valves having the following sets of numbers:

Process Valves 001 to 499

Drain and Vent Valves 500 to 699

Instrument Valves 700 to 999

1.4.5 Instrument Numbering

Equipment numbers for instruments are assigned in a series, for instruments only, from the upstream side of the system. They are assigned without relation to the symbols for the type of equipment; that is, without regard to the variables measured and measuring functions. The following sets of numbers are used for instruments according to their location and equipment classification:

001 to 299 Instruments installed in local panels.

301 to 399 Instruments installed locally, attached to equipment only.

601 to 999 Instruments installed in main control room, including instrument functions performed by multiplexer. The instrument number assigned to the latter is prefixed by the letter Z.

For a system having more than one fluid stream, instruments are numbered in sequence with those used for water first, then for steam and then for air. Within any of the above categories, for local instruments mounted on equipment, the priority is for level instruments first, then pressure and then temperature. Instruments measuring the same quantities, in this case, are numbered in sequence from those which have higher setting values, or from those which have higher upper limit values.

Water

Steam

Air

Locally Mounted*

Rack Mounted

Control Room

(Same as Above)

(Same as Above)

Instruments

1.7-16 Drawings

Rev. 0


1.4.6 Numbering Within a Loop

Instruments or instrument functions performed by the multiplexer in the same loop have the same last two numbers. Instruments located in the main control room that receive signals from locally installed instruments in the same loop are numbered by adding 600 to the local or local panel-mounted instrument number in the loop.

1.5 Equipment Suffix #1

Equipment suffix #1 shall consist of a single letter (A, B, C, etc.). This is assigned when equipment or instruments in a system have the same equipment numbers and are required to be differentiated because of safety/separation considerations or because there are redundant instrument or mechanical loops. The following set of guidelines is followed in assigning equipment suffix #1:

(a) Equipment suffix #1 is the same as the suffix assigned to the reactor vessel nozzle to which the associated system or subsystem containing that equipment or instrument is connected.

(b) For equipment or instruments arranged in parallel and their systems having the same flow direction, the equipment suffix #1 is assigned by the equipment arrangement. The numbering is done from north-to-south or from sea-to-mountain. The north-to-south direction takes priority over the sea-to-mountain direction.

(c) For equipment or instruments having the same flow direction but which are installed at upper and lower levels, the suffix #1 is assigned as “A”, “B”, “C”, etc., from the lower level up.

(d) For the secondary system, the equipment suffix #1 is the same as that for the equipment in the primary system.

(e) If equipment or instruments belonging to an interfacing system are connected to equipment with suffix A and B in the primary system, their equipment suffix #1 may be omitted if they can be differentiated without it.


Level Set Point Value

High

(Same as Above)

(Same as Above)

*Local InstrumentsMounted on Equipment (Special Cases)

Instruments

PressureInstruments

TemperatureInstruments

or RangeLow

Drawings 1.7-17

Rev. 0


(f) When components connected to a dual system are further divided, the equipment suffix #1 is assigned in a staggered fashion. That is, component elements of the secondary system which are connected to system A have suffix A, C, E, G, J, L,...., while those which are connected to system B have suffix as B, D, F, H, K, M.....

(g) The Hydraulic Control Units (HCUs) in the Control Rod Drive (CRD) System shall be assigned a different type of equipment suffix #1. The core-coordinates of the two fuel bundles to which a particular HCU belongs shall be used as suffix. For example, C12D0010722/2718 represents an HCU for control rods belonging to fuel bundles at core coordinates 07,22 and 27,18.

1.6 Equipment Suffix #2

Equipment suffix #2 is only used for instruments if necessary. This number will differentiate instruments of the same type in an instrument loop. A single digit number is used in specifying the equipment suffix #2.


FT001

FE001

FT001

FE001

FT001

FT001

B

B

A

C AD

SYSTEM B SYSTEM A

F101A

F102A

F101C

F102CF102B

F101BF101D

F102D

LS

001

LS

001

A-1

A-2

LT

EQUIPMENT SUFFIX #2

1.7-18 Drawings

Rev. 0


1.7 Numbering of Shared Equipment

The following rules are followed in assigning numbers to equipment shared between systems or between loops within a system:

(a) “System group and number” of shared instruments—Assign the system number of whichever system has the largest number of instruments using the shared component. If the number of instruments is the same, use the system number which has the system group and system number closest to A00.

(b) “Equipment Number” of shared instruments—Except for instruments with a recording function, the same rule as outlined for unshared instruments is followed. All recorders, regardless of the “measured variable”, are numbered in a single series from 001 to 999.

(c) In instrumentation systems which monitor the process quantities of one system and perform interlock controls with another system, the primary instruments (elements and transmitters, or local switches) are assigned the system and equipment number of the system being monitored; and the other instruments are assigned the system and equipment number of the latter system. However, switch functions sending signals to multiple systems are excepted from the above rule and are considered as a part of the primary system.

2.0 Piping and Instrument Diagram Standards

2.1 P&ID

The P&ID provides a schematic illustration of a specific system. It may contain the following information:

(a) Equipment, valves, piping and instrumentation required for system function.

(b) Interface between components and other systems to show control and function of each valve.

(c) Electrical and instrumental interlocks, protective features and logic connections.

(d) Valves and associated components shown in plant normal operating mode (e.g., valve open-valve closed) or as defined on the drawing or specified in the notes. An exception to this is a three-way solenoid valve supplied with associated air or nitrogen-operated valve, which is shown in the de-energized mode.


LT LSRPS

C71B21

Drawings 1.7-19

Rev. 0


(e) The P&ID shows the location of the valves, pipe junctions, pumps, instruments, tanks and other equipment in actual sequence along the pipeline. Piping takeoff connections from equipment are shown at their proper locations relative to the equipment whenever practicable.

(f) System design (maximum) conditions such as design pressure, design temperature, material and seismic class are given in the P&ID. The changing points for these items are defined.

(g) The identification of building(s) (including yard) is defined.

(h) Equipment, valves and instrumentation belonging to another system or used in common are shown by broken line with two dots between each line break, and the system group and number(s) clearly stated for other system or systems.

(i) Instrument root valves in the instrument piping branching from the process piping are shown. Valves on the instrument side are not shown.

(j) Drain, vent and test connections are shown on P&IDs. The discharge of drains and vents is assigned to the appropriate drain system whose system acronym is written at the end of the line.

(k) System (group and number) and system acronyms are given in the upper right-hand corner of the first sheet of the P&ID.

(l) Use of a black box on a P&ID is allowed when other sheets of the same drawing or a different drawing contains complete information about the contents of the black box. A note is added that specifies the drawing number of the contents of the black box.

(m) Piping is divided into three categories—Process piping, drain and vent piping and instrument piping. The following sets of numbers are used for these categories:

Process Piping 001 to 499

Drain and Vent Piping 500 to 699

Instrument Piping 700 to 999

When numbers in a series run out, four-digit pipe numbers may be used. For example, for process piping, after 499, the numbers from 1001 to 1499 are used.

(n) The pipe numbering is done using the flow direction method, same as the equipment numbering method described in Subsection 1.4.1.

(p) Piping is basically identified by a single number (Example 1 below). If the P&ID is changed during the detailed design after the initial numbering and if an additional pipe number is required due to the change, a suffix number may be applied (Example 2):

Example 1 400A-MUWC-001

Example 2 400A-MUWC-001-1

400A-MUWC-001-2

(q) Nominal pipe diameter is identified by the symbol “A” preceeded by a millimeter dimension consistent with inches or the ANSI standard outside diameter as shown by the following examples:


1.7-20 Drawings

Rev. 0



Nominal ABWR P&ID ANSIDiameter Symbol OD Inches mm mm

2 50A 60.34 100A 114.38 200A 219.1

16 400A 406.424 600A 609.6

3.0 Process Flow Diagram Standards

3.1 PFD

The process flow diagram shows the engineering requirements or conditions (e.g., modes of operation, flow, pressure and temperature) at specified locations throughout the system using the following guidelines:

(a) Main flow lines of the system are shown. Drain lines, vent lines and instrument lines are not shown.

(b) Identification numbers of the main valves are included. All symbols used are the same as the P&ID.

(c) Operating conditions for each mode of operation are shown in a tabular form.

(d) The position nodes for the key locations at which the operating conditions are given are shown by the symbol (circle) or (hexagon).

3.6 Operating Conditions

The operating conditions include the following items:

(a) Flow (m3/h)

(b) Pressure (kPaG or kPaA)

(c) Temperature (°C)

(d) Valve opening/closing conditions.

(e) Maximum pressure drop (m) if necessary.

Drawings 1.7-21

Rev. 0


The following figure is located in Chapter 21 :

Figure 1.7-1 Piping and Instrumentation Diagram Symbols (Sheets 1–2)

1.7-22 Drawings

Rev. 0


Figure 1.7-2 Graphical Symbols for Use in IBDs (Sheet 1 of 8)

No. Function Graphic Symbol Explanation of Function

1 Condition Symbol or

Signal

Symbol indicates signal condition or action (e.g. valve close signal). Action shifts to right when condition is met.Y – Is instrument numberZ – Represents the name of condition signalS – Shows above or below setpoint for

transfer of signal condition

2A AND Output exists if and only if all specified inputs exist

2B

2C 4-Input AND Truth Table (Not Shown)

Y X

Y XOR

Z S

CAB

2-Input AND

A B C0 0 00 0

01

11

01 1

2-Input AND Truth Table

3-Input AND

DB

C

A A B C0 0 00 0

01

11

D0000000

11 1 11 1

110

00 01

00 1

3-Input AND Truth Table

4-Input AND

EB

D

A

C

Drawings 1.7-23

Rev. 0


2D Coincident Variable Gate

(CVG)

Output exists if specified number of inputs exist (2 of 3, 2 of 4, or 3 of 4)

2E

2F 2/4 and 3/4 AND’s Truth Tables (Not Shown)

3A

3B

OR Output exists only when at least one input exist



2/3 AND

DB

C

A

2/3A B C0 0 00 0

01

00

D0000111

11 1 11 1

110

01 10

10 1

Truth Table 2/3 AND

2/4 AND

EB

D

A

C2/4

3/4 AND

EB

D

A

C3/4

CA

B

2-Input OR

D

A

B

3-Input OR

C

A B C0 0 00 1

11

11

01 1

Truth Table 2-Input OR

A B C0 0 00 0

01

00

D0111111

11 1 11 1

110

01 10

10 1

Truth Table 3-Input OR

1.7-24 Drawings

Rev. 0


3C X OR (Exclusive “OR”)

These logic symbols represent an Exclusive “OR” Gate whose output assumes 1 state if one and only one of the logic input assumes the 1 state

Truth Table 2-Input Exclusive “OR”

3D


3E Exclusive“OR”




CA

B

2-Input X OR

A B C0 0 00 1

11

10

01 1

D

A

B

3-Input X OR

C

A B C0 0 00 0

01

00

D0111000

11 1 01 1

100

10 11

10 1

E

AB

4-Input X OR

DC

A B C

0 0 00 0

10

01

D

0100011

01 0 10 1

100

00 01

00 0

E

01111000

0 1 11 0

01

11

0001110

11 1 11 1

011

10 11

01 1

00000000

Drawings 1.7-25

Rev. 0


4 Not This symbol shows the “NOT” condition. Output B is opposite to input A

5A Timer Elements TPU– Signal B is energized within specified time limit (t) after signal A is energized. B terminates when A terminates.

5B TDO– Initially B is energized when A is energized. signal B is de-energized within specified time limit

(t) after signal A is de-energized

6A Wipe-Out (Signal Block)

When signal C is not present, signal A is transmitted to B. When signal C is present, signal A is stopped and does not flow to B. (WO: Wipe-out)

6B Delayed Wipe-Out

(One-Shot)The output signal to B is stopped after time interval “t”.



BAA B0 11 0

A B

Delayed

t sec

Initiation

TPU

tA

B

A B

Delayed

t sec

Termination

TDO

(Reset) t

A

B

(WO)A B

C

A C B0 0 00 1

10

0111 0

t sec

TPU(WO)A B

t

A

B

1.7-26 Drawings

Rev. 0


7A Self-Hold or Reset

When condition C does not exist, Condition A holds itself and there is output to B. The self holding is released when condition C is established and there is an output to B only when there is an A condition (A takes priority).

7B When condition C does not exist, Condition A holds itself and there is output to B. The self holding is released when condition C is established and there is no output to B (C takes priority).

8 Operating Switch

S – Place of installationX – Switch operation nameY – Switch type, e.g.

CS—Control Switch Spring ReturnCOS—Control Operating Switch PositionHoldPBS—Pushbutton SwitchPBL—Pushbutton Illuminated TypeKS—Key Switch (Spring Return)KOS—Key Operating Switch (PositionHold)CRT—CRT Touch-Screen

Z – Switch Position: On, Off, Pull Hold, … etc.

9 Control Component or

Device

Shows a component or device to be controlledX – Part # of controlled deviceY – Controlled device name e.g. pump,

valve, etc.Z – Controlled condition, e.g., Start, Stop, On,

Off, Open, Close, … etc.

10 Electromagnetic Valve

This symbol represents an electromagnetic valveE – EnergizeDE – De-energized



BA

(WO)

C

BA

(WO)

C

XZZZ

Y

S

ZZ

Y

ZZ

XY

Fully OpenFully Close

EDE

Drawings 1.7-27

Rev. 0


11 Electromagnetic PilotValve

This symbol represents an electromagnetic pilot valveE – EnergizedDE – De-energized

12 Memory (Flip-Flop)

S Represents “Set Memory”R Represents “Reset Memory”

Logic output C exists when logic input A exists.C continues to exist regardless of subsequent state of A and until reset by input at B

C remains terminated regardless of subsequent state of B, until A causes memory to reset. Logic output D, if used, exist when C does not exist, and D does not exist when C exists.

13 Static Transducer (Converter)

This device converts “E” (Electrical Signal) to “P” (Pneumatic Signal)

14 Electromagnetic Pilot Valvefor Control

Shows a pilot electromagnetic valve for a control valve. When the pilot electromagnetic valve is energized by a signal from A, opening of the control valve is adjusted by a signal from B.

15 TransmissionSignals or Lines Indicates electrical signal and flow direction

Indicates pneumatic line and flow direction

Indicated oil hydraulic pressure line and flow directionIndicates mechanical linkage

16 Electrical Signal

ConnectionSignal is connected electrically

Signal is not connected electrically

17 Signal Input This symbol represents an input signal to a computer, display, test panel, etc. as designated by the letter X inside the triangle. The letter N indicates the assigned signal number.

18 Operational Condition

This graphical presentation in used in sequential control



EDE

SR

AB

CD*

*Output D shall not beshown if not used.

E/P

E/P

EDE

B

A

XN

“A” ValveFully Open

1.7-28 Drawings

Rev. 0


19 Virtual Condition

Signal

Used only for signals which do not actually exist but are convenient to show on IBD.

20 PanelIndicator Lights Red indication light:

Shows actuation, input and valve opening

Green indication light:Shows stop, interruption and valve closure

White (milk-white) indication light:Shows condition indication, automatic mode operation… etc.

Orange indication light:Shows caution and failure

Colorless or transparent indicating light

CRT Indicator Lights Light indicator to be shown on CRT. X represents

the color of the light to be indicated.

Alarm Indicates an annunciated alarm or warning. The letter N indicates the alarm number.

21 Isolator This symbol represents that the input signal shall be divisionally isolated from the output signalX – Input division numberY – Isolation output division number

22 to 32 Intentionally left blank for future additions

33 Signal Transfer These symbols indicate signal transfer to other location(s). The upper half of the symbol is used to enter the transfer code. The lower left portion of symbol is used to reference the sheet number to go to, and the lower right hand portion will indicate the location where the signal can be found. The transfer code shall utilize either an English letter or a number if signal transfer is within the same sheet or to other sheets of the IBD. For signal transfer from/or to other MPL systems, the transfer code shall be expressed with 2 English letters starting with “AA”. Also indicate the system MPL reference where the signal goes to or originates next to the symbol.



R

G

W

O

T

X CRT

A

N

X

Y

OR

Drawings 1.7-29

Rev. 0


34MOV’s Position

IndicationDesignators

(See Appendix “D” for

ApplicationExamples)

LS: Limit SWTS: Torque SW

The above letter designators are used to show the control methods of motor-driven valves. The control method should be indicated above the left side of the component block for valve “Opening” and below the left side for valve “Closure”.

35A ComparatorThese symbols represent a comparator that provide an output when the condition A ≥ B or A≤ B is met.

35B

36A Load Driver

This symbol represents a standard load driver.

36B

This Symbol represents a load driver whose output power signal DIV 3 is isolated internally from the input logic signal division 1.



(LS)L

MOVOPEN

CLOSE

TL(TS) (LS)

LL(T)TLTTL(C)T(C)

— Limit Off— Limit Off With Torque Backup— Both Limit and Torque Off— Torque Off— With Chattering Prevention at TL— With Chattering Prevention at T

A ≥ B

A ≤ B

AB

AB

EDE

AC or DC

LoadDriver

ACorDC

StandardLoad Driver

EDE

AC or DC

LoadDriver

ACorDC

IsolatedLoad Driver

1

3

1.7-30 Drawings

Rev. 0


Figure 1.7-3 Graphical Symbols for Use in Electrical SLDs (Sheet 1 of 4)

Transformer

Three 1ø

Three winding - 3ø

Two winding - 3ø or 1ø

Two winding - 3ø

Two winding - 3ø center tap ground

Note: Symbol in transformer = winding connection type i.e.

Star

Delta

Special

Drawings 1.7-31

Rev. 0


Circuit Breakers/Disconnect Devices

Main Generator/Switchyard Circuit Breaker

Metal Clad Switchgear Circuit Breaker (Drawout Type)

Air Circuit Breaker (Drawout Type)

Isolated Phase Bus Disconnect Link

Disconnect Switch

Generators

Main Plant Generator

Diesel Generator


MAINGEN

DG

1.7-32 Drawings

Rev. 0


Combustion Turbine Generator

Motor Generator

Ground

M/C, P/C Switchgear Ground

Transformer Ground

Overload/Protection Devices

Magnetic Overload Devices

Thermal Overload Devices

Electrical Protection Assembly


CTG

M G

EPA

Drawings 1.7-33

Rev. 0


Fuse

DC Supply

Station Battery

Full Wave Rectifier

Devices

Current Transformer (CT)

*Function IdentifierSTD - Static Trip Device27D - DC under voltage relay64 - Ground detector relay76 - DC overcurrent relay84 - Selector switch

* Monitoring FunctionV - VoltmeterA - Ampmeter

Abbreviations

M/C Metal CladP/C Power CenterMCC Motor Control CenterR/B Reactor BuildingC/B Control BuildingT/B Turbine BuildingRW/B Radwaste BuildingUHS Ultimate Heat Sink


*

*

/34

Conformance with Standard Review Plan and Applicability of Codes and Standards 1.8-1

Rev. 0


1.8 Conformance with Standard Review Plan and Applicability of Codes and Standards

1.8.1 Conformance with Standard Review Plan

This subsection provides the information required by 10CFR50.34(g) showing conformance with the Standard Review Plan (SRP). The summary of differences from the SRP section is presented by SRP section in Tables 1.8-1 through 1.8-18. (See Subsection 1.8.4.1 for COL license information.)

1.8.2 Applicability of Codes and Standards

Standard Review Plans, Branch Technical Positions, Regulatory Guides and Industrial Codes and Standards which are applicable to the ABWR design are provided in Tables 1.8-19, 1.8-20 and 1.8-21. Applicable revisions are also shown.

1.8.3 Applicability of Experience Information

Experience information is routinely made available and distributed to design personnel in the design process. Nuclear field experience is maintained in hard copy form in functional component and library files and in the GE world-wide computer retrieval system.

Generic Letters and IE Bulletins, Information Notices and Circulars covering the decade including 1980 through the current issues (late 1991) were reviewed for applicability to the ABWR design. The review was enhanced by associating related experiences and tracing referenced occurrences. This was accomplished starting with the current issues of the Generic Letters and proceeding back into the decade. The Circulars, Bulletins and Notices were reviewed in that order. Interfacing experience was included in the review. The selection of ABWR information was based on the significance to future design and operation guidance. Included is a list of NUREGs related to the closing of current safety issues. Experience that resulted in applicable rules, codes and standards was not repeated. Table 1.8-22 lists the experience information that has been included in the ABWR design or impacts the COL applicant. (See Subsection 1.8.4.2 for COL license information.)

A systematic procedure encompassing available resources was used to identify the applicability of experience information resulting in Table 1.8-22. Engineering management surveyed the indices of annual experience information to identify those very likely to be applicable to the ABWR. The remaining potentially applicable experiences were reviewed individually. Experience information not deemed applicable to the ABWR design (issues pertaining to other reactor types, scram discharge volume, etc.) were not included in Table 1.8-22. The experience information categories applicable to the ABWR design in Table 1.8-22 include experience information accommodated by a design change, covered by review of USIs/GSIs or an

1.8-2 Conformance with Standard Review Plan and Applicability of Codes and Standards

Rev. 0


issue that impacts the ABWR design but must be addressed by the COL applicant. This latter category is included as COL license information.

Experiences related to identified regulatory or industry developed resolutions were eliminated to avoid repetition except for selected experiences that have a nuisance potential for reoccurring. Lead system engineers classified the more complex experiences.

Reference to the new or novel design features used in the ABWR are provided below:


1.8.4.1 SRP Deviations

The SRP sections to be addressed by the COL applicant are indicated in the comments column of Table 1.8-19 as “COL Applicant”. Where applicable the COL applicant will provide the information required by 10CFR50.34(g) similar to Tables 1.8-1 through 1.8-18 (see Subsection 1.8.1).

1.8.4.2 Experience Information

The experience information to be addressed by the COL applicant are indicated in the comment column of Table 1.8-22 as “COL Applicant” (see Subsection 1.8.3).

Feature Tier 2 Section

Fine Motion Control Rod Drive 4.6

Internal Reactor Pumps 5.4.1

Multiplexing 7A.2

Digital/Solid-State Control 7A.7

Overpressure Protection System 6.2.5.2.6,6.2.5.3,6.2.5.4

AC-Independent Water Addition System 5.4.7.1.1.10

Lower Drywell Flooder 9.5.1.2


Rev. 0


Table 1.8-1 Summary of Differences from SRP Section 1

SRP Section

Specific SRP

Acceptance Criteria

Summary Description of

Difference

Subsection Where

Discussed

None None None None


SRP

Section

Specific SRP

Acceptance Criteria


Difference

Subsection

Where Discussed

2.2.1-2.2.2

See Table 2.1-1. Limits imposed on selected SRP Section II acceptance criteria by (1) the envelope of the ABWR Standard Plant site parameters and (2) evaluations assumptions.

2.1

2.2.3

2.3.1

2.3.4

2.4.1

2.4.4

2.4.5

2.4.6

2.4.8

2.4.11.6

2.4.12

2.5.2.7 OBE is not a design requirement.


SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection

Where Discussed

3.6.1 and 3.6.2

II—Postulated pipe rupture.

Large bore piping can utilize leak before break option as provided in GDC-4 October 27, 1987, “Modification of General Design Criterion 4”.

3.6 and 3.6.3

3.6.1 and 3.6.2

ASB 3-1 and MEB 3-1 Consider 1/2 SSE for postulating pipe ruptures.

Earthquake stresses considered only in cumulative usage factor calculations when postulating pipe ruptures

3.6.1.1, 3.6.2.1.4.2, 3.6.2.1.4.3,3.6.2.1.4.4.,3.6.2.1.5.2,3.6.2.1.5.3


Rev. 0


3.6.2 MEB 3-1, B.1.c.(1).(b) - Pipe ruptures must be postulated if Eq.(10) of NB-3653 exceeds 2.4 Sm.

Pipe ruptures postulated only if, in NB-3653, Eq. (10) and either (12) or Eq. (13) exceed 2.4 Sm.

3.6.1.1 and 3.6.2.1.4.3

3.7.1 and3.7.3

II.2 - Two earthquakes, the SSE and the OBE shall be considered in the design.

The ABWR will be based on a single earthquake (SSE) design.

3.6, 3.7, 3.9

3.9.2 II-E.2.g - For multiply supported equipment use envelope RS and;

Independent Support Motion Response Spectrum methods acceptable for use.

3.7.3.8.1.10

Combine responses from inertia effects with anchor displacements by Absolute Sum.

Combine responses from inertia effects with anchor displacements by SRSS.

3.7.3.8.1.8

3.7.3 II.2.b—Determination of number of OBE cycles.

The ABWR is based on a single earthquake (SSE) design, two SSE events with 10 peak stress cycles per event are used.

3.7.3.2


SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection

Where Discussed

None None None None


SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection

Where Discussed

5.2.3 II.3.b.(3)—Reg Guide 1.71, Welding Qualification for Areas of Limited Accessibility.

Alternate position employed. 5.2.3.4.2.3

5.2.4 II.1—Inspection of Class 1 pressure-containing components.

Some welds inaccessible for volumetric examination.

5.2.4.2.2

5.4.6 Deleted

Table 1.8-3 Summary of Differences from SRP Section 3 (Continued)

SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection

Where Discussed


Rev. 0


5.4.7 Branch Technical Position RSB 5-1, B.1.(b) and (c)—Diverse interlocks for RHR suction isolation valves.

No diversity of interlocks. 5.4.7.1.1.7


SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection

Where Discussed

6.2.1.1 Design provision for automatic actuation of wetwell spray 10 minutes following a LOCA signal

Manual actuation of wetwell spray 30 minutes following a LOCA signal

6.2.1.1.5.6.1

6.2.4 One isolation valve inside and one isolation valve outside containment

Both isolation valves located outside the containment

6.2.4.3.2.2.2.3

Purge and vent valves to close in

≤5 secondsPurge and vent valves will close in

≤20 seconds6.2.4.3.2.2.2.3

6.2.1.1 Monthly vacuum valve operability test

Operability tests only performed during refueling outages

6.2.1.1.5.6.3


SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection

Where Discussed

7.1 Table 7-1: 1a IEEE-279, 4.19

RHR Annunciation at loop level. 7.3.2.3.2 (1) 7.3.2.4.2 (1) 7.4.2.3.2 (1)

7.1 Table 7-1: 2i GDC 20

Some modes of RHR are not automatic.

7.3.2.3.2 (2)(b) 7.3.2.4.2 (2)(b) 7.4.2.3.2 (1)

7.1 Table 7-1: 3a Reg Guide 1.22

Clarification of requirements. 7.3.2.1.2. (3)(a)

7.1 Table 7-1: 3a Reg Guide 1.22

HP/LP interlocks cannot be tested during power operation.

7.6.2.3.2 (3)

7.1 Table 7-1: 3c Reg Guide 1.53

Continuity testing of certain solenoids.

7.3.2.1.2. (3)(c)


Some components are not redundant.

7.3.2.5.2 (3)7.4.2.2.2 (3)


SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection

Where Discussed


Rev. 0



Limited redundancy of remote shutdown.

7.4.2.4.2 (1) 7.4.2.4.2 (3)

7.1 Table 7-1: 3e Reg Guide 1.75

Alternate positions employed. 7.1.2.10.5

7.1 Table 7-1: 3h Reg Guide 1.118

Some sensors cannot be tested at power operation.

7.2.2.2.1 (7) 7.2.2.2.3.1 (10) 7.2.2.2.3.1 (21)

7.1 Table 7-1: 4i BTP ICSB 22

Some actuators cannot be exercised during power operation.

7.3.2.1.2 (4)(d) 7.4.2.3.2 (4)(c)


SRP Section

Specific SRP

Acceptance Criteria Summary Description of Difference

Subsection

Where Discussed

8.1 Table 8-1: 2f Reg Guide 1.75

Exception to LOCA trip for certain non-1E loads.

8.1.3.1.2.2 (6) Appendix 9A

8.1 Table 8-1: 2f Reg Guide 1.75

4.572 m cable marking intervals. 8.3.3.5.1.3

8.1 Table 8-1: 2f IEEE-384

LDS divisional separation in steam tunnel.

8.3.3.6.1.2 (2)


SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection

Where Discussed

9.3.1 II.1—Particles shall not exceed 3 micrometer.

Instrument air is filtered to 5 micrometer.

9.3.6.2

9.3.2 II.k.5—Capable of sampling liquid of 370,000 MBq/cm3.

Capable of sampling liquids of 37,000 MBq/cm3.

9.3.2.3.1

9.4.1 GDC 19 Site specific. 6.4.7.1

9.5.1 Section 7.b Control Room Complex1. Peripheral rooms2. Underfloor (subfloor)3. Consoles & cabinets

9.5.1

Section 7.j Diesel fuel oil tank capacity 9.5.1

Section 7.i Diesel Generator operation 9.5.1


SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection

Where Discussed


Rev. 0



SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection Where

Discussed

None None None None


SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection Where

Discussed

11.1 II.9—BWR GALE Code Alternate computer code. 20.3.7 (Response to Question 460.1)


SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection Where

Discussed

None None None None


SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection Where

Discussed

None None None None


SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection Where

Discussed

None None None None


SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection Where

Discussed

15.1.1-15.1.4

Acceptable analytical model.

3D simulator instead of REDY Code is used. 3D simulator was approved for use in GESTAR review by NRC.

15.1.1.3.2


Rev. 0


15.2.6 All recirculation pumps are tripped simultaneously by the initiating event.

Only four of ten RIPs are tripped. This is based on ABWR design.

15.2.6.1.1

15.3.1-15.3.2

Complete recirculation pumps trip is considered as a moderate-frequencytransient.

Trip of all RIPs is classified as an infrequent low probability event with special acceptance for fuel failure.

15.3.1.1.2

15.3.3-15.3.4

II.10—coincidentturbine trip, loss of offsite power and coastdown of undamaged pumps.

Not analyzed with the assumption. If the assumption is made, the consequence would be similar to the event shown in 15.2.6.

15.3.3.2.2.

15.4.2 Analysis of uncontrolled control rod withdrawal at power.

No quantitative analysis is provided because ABWR’s ATLM design prevents this transient from occurring.

15.4.2.2

15.4.4-15.4.5

II.2.(b)—Fuel cladding integrity.

MCPR not calculated, since transients are very mild.

15.4.4.315.4.5.3.2.1 and 15.4.5.3.2.2

15.4.9 Not applicable SRP for BWR.

Discussion is provided to show this event cannot occur with ABWR FMCRD design.

15.4.9

15.4.10 Analysis of rod drop accidents.

No quantitative analysis is provided because ABWR’s FMCRD design prevents this accident from occurring.

15.4.10.1 & 15.4.10.2

15.6.5 II.2—Use of assumptions outlined in Reg Guide 1.3.

ABWR LOCA analysis incorporates suppression pool scrubbing IAW SRP 6.5.5 and in variance from R.G. 1.3. Fission product plateout and removal is incorporated in the analysis of leakage sources through the main steamlines and into the turbine condenser based upon BWROG analysis of acceptability of the steamlines and condenser to mitigate releases without requiring Seismic Category I structures.

15.6.5.5


SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection Where

Discussed


Rev. 0



SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection Where

Discussed

None None None None


SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection Where

Discussed

17.1 II.1 - Applicant is responsible for overall QA program.

GE & major technical associates are responsible for their own QA programs.

17.017.1.117.1.2

17.1 II.2,3,4,7,13,17,18-Meet identified quality related Reg Guides.

Reg Guide 1.28, Rev. 3 and alternative positions employed.

Table 17.0-1 17.1.2, 17.1.3 17.1.4, 17.1.7 17.1.13, 17.1.17 17.1.18

17.1 II.2 - Meet identified regulations and codes.

Differences between domestic and international designs are identified in a controlled list.

17.1.3


SRP Section

Specific SRP

Acceptance Criteria


Difference

Subsection Where

Discussed

None None None None


Rev. 0


Table 1.8-19 Standard Review Plans and Branch Technical Positions

Applicable to ABWR

SRP No. SRP Title

Appl.

Rev.

Issued

Date

ABWR

Appli-

cable? Comments

Chapter 1 Introduction and General Description of Plant

1.8 Interfaces for Standard Design 1 7/81 Yes

Chapter 2 Site Characteristics

2.1.1 Site Location and Description 2 7/81 — COL Applicant

2.1.2 Exclusion Area Authority and Control 2 7/81 — COL Applicant

2.1.3 Population Distribution 2 7/81 — COL Applicant

2.2.1–2.2.2

Identification of Potential Hazards in Site Vicinity 2 7/81 — COL Applicant

2.2.3 Evaluation of Potential Accidents 2 7/81 — COL Applicant

2.3.1 Regional Climatology 2 7/81 — COL Applicant

2.3.2 Local Meteorology 2 7/81 — COL Applicant

2.3.3 Onsite Meteorological Measurements Programs 2 7/81 — COL Applicant

Appendix A 2 7/81 — COL Applicant

2.3.4 Short-Term Diffusion Estimates for Accidental Atmospheric Releases

1 7/81 — COL Applicant

2.3.5 Long-Term Diffusion Estimates 2 7/81 — COL Applicant

2.4.1 Hydrologic Description 2 7/81 — COL Applicant


2.4.2 Floods 2 7/81 — COL Applicant

2.4.3 Probable Maximum Flood (PMF) on Streams and Rivers



Rev. 0


2.4.4 Potential Dam Failures 2 7/81 — COL Applicant

2.4.5 Probable Maximum Surge and Seiche Flooding 2 7/81 — COL Applicant

2.4.6 Probable Maximum Tsunami Flooding 2 7/81 — COL Applicant

2.4.7 Ice Effects 2 7/81 — COL Applicant

2.4.8 Cooling Water Canals and Reservoirs 2 7/81 — COL Applicant

2.4.9 Channel Diversions 2 7/81 — COL Applicant

2.4.10 Flood Protection Requirements 2 7/81 — COL Applicant

2.4.11 Cooling Water Supply 2 7/81 — COL Applicant

2.4.12 Groundwater 2 7/81 — COL Applicant

BTP HGEB 1 2 7/81 — COL Applicant

2.4.13 Accidental Releases of Liquid Effluents in Ground and Surface Waters


2.4.14 Technical Specifications and Emergency Operation Requirements


2.5.1 Basic Geologic and Seismic Information 2 7/81 — COL Applicant

2.5.2 Vibratory Ground Motion 2 8/89 — COL Applicant

2.5.3 Surface Faulting 2 7/81 — COL Applicant

2.5.4 Stability of Subsurface Materials and Foundations 2 7/81 — COL Applicant

2.5.5 Stability of Slopes 2 7/81 — COL Applicant

Chapter 3 Design of Structures, Components, Equipment, and Systems

3.2.1 Seismic Classification 1 7/81 Yes


Applicable to ABWR (Continued)

SRP No. SRP Title

Appl.

Rev.

Issued

Date

ABWR

Appli-

cable? Comments


Rev. 0


3.2.2 System Quality Group Classification 1 7/81 Yes

Appendix A (Formerly BTP RSB 3-1) 1 7/81 Yes

Appendix B (Formerly BTP RSB 3-2) 1 7/81 Yes

3.3.1 Wind Loadings 2 7/81 Yes

3.3.2 Tornado Loadings 2 7/81 Yes

3.4.1 Flood Protection 2 7/81 Yes

3.4.2 Analysis Procedures 2 7/81 Yes

3.5.1.1 Internally Generated Missiles (Outside Containment)

2 7/81 Yes

3.5.1.2 Internally Generated Missiles (Inside Containment) 2 7/81 Yes

3.5.1.3 Turbine Missiles 2 7/81 Yes

3.5.1.4 Missiles Generated by Natural Phenomena 2 7/81 Yes

3.5.1.5 Site Proximity Missiles (Except Aircraft) 1 7/81 Yes

3.5.1.6 Aircraft Hazards 2 7/81 Yes

3.5.2 Structures, Systems, and Components to be Protected from Externally Generated Missiles

2 7/81 Yes

3.5.3 Barrier Design Procedures 1 7/81 Yes

[Appendix A 0 7/81](1) Yes

3.6.1 Plant Design for Protection Against Postulated Piping Failures in Fluid Systems Outside Containment

1 7/81 Yes

BTP ASB-3-1 1 7/81 Yes

3.6.2 Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping

1 7/81 Yes

BTP MEB-3-1 2 6/87 Yes

3.7.1 Seismic Design Parameters 2 8/89 Yes

3.7.2 Seismic System Analysis 2 8/89 Yes

3.7.3 Seismic Subsystem Analysis 2 8/89 Yes

3.7.4 Seismic Instrumentation 1 7/81 Yes

3.8.1 Concrete Containment 1 7/81 Yes

[Appendix 0 7/81](1) Yes



SRP No. SRP Title

Appl.

Rev.

Issued

Date

ABWR

Appli-

cable? Comments


Rev. 0


3.8.2 Steel Containment 1 7/81 Yes

3.8.3 Concrete and Steel Internal Structures of Steel or Concrete Containments

1 7/81 Yes

3.8.4 Other Seismic category I Structures 1 7/81 Yes

Appendix A 0 7/81 Yes

Appendix B 0 7/81 Yes

Appendix C 0 7/81 Yes

Appendix D 0 7/81 Yes

3.8.5 Foundations 1 7/81 Yes

3.9.1 Special Topics for Mechanical Components 2 7/81 Yes

3.9.2 Dynamic Testing and Analysis of Systems, Components, and Equipment

2 7/81 Yes

3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures

1 7/81 Yes


3.9.4 Control Rod Drive Systems 2 4/84 Yes

3.9.5 Reactor Pressure Vessel Internals 2 7/81 Yes

3.9.6 Inservice Testing of Pumps and Valves 2 7/81 Yes

3.10 Seismic Qualification of Category I Instrumentation and Electrical Equipment

2 7/81 Yes

3.11 Environmental Design of Mech. and Elec. Equip. 2 7/81 Yes

Chapter 4 Reactor

4.2 Fuel System Design 2 7/81 Yes

[Appendix A 0 7/81](2) Yes

4.3 Nuclear Design 2 7/81 Yes

BTP CPB 4.3-1 2 7/81 Yes

4.4 Thermal and Hydraulic Design 1 7/81 Yes

4.5.1 Control Rod Drive Structural Materials 2 7/81 Yes

4.5.2 Reactor Internal and Core Support Materials 2 7/81 Yes

4.6 Functional Design of Control Rod Drive System 1 7/81 Yes



SRP No. SRP Title

Appl.

Rev.

Issued

Date

ABWR

Appli-

cable? Comments


Rev. 0


Chapter 5 Reactor Coolant System and Connected Systems

5.2.1.1 Compliance with the codes and Standard Rule, 10CFR50.55a

2 7/81 Yes

5.2.1.2 Applicable Code Cases 2 7/81 Yes

5.2.2 Overpressure Protection 2 7/81 Yes

BTP RSB 5-2 0 7/81 No PWR only

5.2.3 Reactor Coolant Pressure Boundary Materials 2 7/81 Yes

BTP MTEB 5-7 (Superseded by NUREG-0313)

5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing

1 7/81 Yes

5.2.5 Reactor Coolant Pressure Boundary Leakage Detection

1 7/81 Yes

5.3.1 Reactor Vessel Materials 1 7/81 Yes

5.3.2 Pressure-Temperature Limits 1 7/81 Yes

BTP MTEB 5-2 1 7/81 Yes

5.3.3 Reactor Vessel Integrity 1 7/81 Yes

5.4 Deleted

5.4.1.1 Pump Flywheel Integrity (PWR) 1 7/81 No PWR only

5.4.2.1 Steam Generator Materials 2 7/81 No PWR only

BTP MTEB 5-3 2 7/81 No

5.4.2.2 Steam Generator Tube Inservice Inspection 1 7/81 No PWR only

5.4.6 Reactor Core Isolation Cooling System (BWR) 3 7/81 Yes

5.4.7 Residual Heat Removal (RHR) System 3 7/81 Yes

BTP RSB 5-1

5.4.8 Reactor Water Cleanup System (BWR) 2 7/81 Yes

5.4.11 Pressurizer Relief Tank 2 7/81 No PWR only

5.4.12 Reactor Coolant System High Point Vents 0 7/81 Yes

Chapter 6 Engineered Safety Features

6.1.1 Engineered Safety Features Materials 2 7/81 Yes

BTP MTEB 6-1 2 7/81 Yes



SRP No. SRP Title

Appl.

Rev.

Issued

Date

ABWR

Appli-

cable? Comments


Rev. 0


6.1.2 Protective Coating Systems (Paints)—Organic Materials

2 7/81 Yes

6.2.1 Containment Functional Design 2 7/81 Yes

6.2.1.1.A PWR Dry Containments, Including Subatmospheric Containments

2 7/81 No PWR only

6.2.1.1.B Ice Condenser Containments 2 7/81 No PWR only

6.2.1.1.C Pressure Suppression Type BWR Containments 6 8/84 Yes



6.2.1.2 Subcompartment Analysis 2 7/81 Yes

6.2.1.3 Mass and Energy Release Analysis for Postulated Loss-of-Coolant Accidents

1 7/81 Yes

6.2.1.4 Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures

1 7/81 Yes

6.2.1.5 Minimum Containment Pressure Analysis for Emergency Core Cooling System Performance Capability Studies

2 7/81 No PWR Only

BTP CSB 6-1 2 7/81 No PWR only

6.2.2 Containment Heat Removal Systems 4 10/85 Yes

6.2.3 Secondary Containment Functional Design 2 7/81 Yes

BTP CSB 6-3 2 7/81 Yes

6.2.4 Containment Isolation System 2 7/81 Yes

BTP CSB 6-4 2 7/81 Yes

6.2.5 Combustible Gas Control in Containment 2 7/81 Yes


BTP CSB 6-2 (Superseded by Reg. Guide 1.7)

6.2.6 Containment Leakage Testing 2 7/81 Yes

6.2.7 Fracture Prevention of Containment Pressure Boundary

0 7/81 Yes

6.3 Emergency Core Cooling System 2 4/84 Yes

BTP RSB 6-1 1 7/81 Yes

6.4 Control Room Habitability Systems 2 7/81 Yes



SRP No. SRP Title

Appl.

Rev.

Issued

Date

ABWR

Appli-

cable? Comments


Rev. 0



6.5.1 ESF Atmosphere Cleanup Systems 2 7/81 Yes

6.5.2 Containment spray as a Fission Product Cleanup System

1 7/81 Yes

6.5.3 Fission Product Control Systems and Structures 2 7/81 Yes

6.5.4 Ice Condenser as a Fission Product Cleanup System

2 7/81 No PWR only

6.5.5 Pressure Suppression Pool as a Fission Product Cleanup System

0 12/88 Yes

6.6 Inservice Inspection of Class 2 and 3 Components 1 7/81 Yes

6.7 Main Steam Isolation Valve Leakage Control System (BWR)

2 7/81 No

Chapter 7 Instrumentation and Controls

7.1 Instrumentation and Controls Introduction 3 2/84 Yes

Table 7-1 Acceptance Criteria and Guidelines for Instrumentation and Controls Systems Important to Safety

3 2/84 Yes

Table 7-2 TMI Action Plan Requirements for Instrumentation and Controls Systems Important to Safety

0 7/81 Yes



7.2 Reactor Trip System 2 7/81 Yes

Appendix A (Superseded by SRP 7.1 App. B)

7.3 Engineered Safety Features Systems 2 7/81 Yes

Appendix A (Superseded by SRP 7.1 App. B)

7.4 Safe Shutdown Systems 2 7/81 Yes

7.5 Information Systems Important to Safety 3 2/84 Yes

7.6 Interlock Systems Important to Safety 2 7/81 Yes

7.7 Control Systems 3 2/84 Yes

Appendix 7-A Branch Technical Positions (ICSB) 2 7/81 Yes

BTP ICSB 1 (DOR) (Deleted)



SRP No. SRP Title

Appl.

Rev.

Issued

Date

ABWR

Appli-

cable? Comments


Rev. 0


BTP ICSB 3 2 7/81 Yes

BTP ICSB 4 (PSB) 2 7/81 Yes

BTP ICSB 5 (Superseded by Std. Tech Specs)





BTP ICSB 16 (Deleted)

BTP ICSB 19 (Deleted)






Appendix 7-B General Agenda, Station Site Visits 1 7/81 Yes

Chapter 8 Electric Power

8.1 Electric Power-Interaction 2 7/81 Yes

Table 8-1 Acceptance Criteria and Guidelines for Electric Power Systems

2 7/81 Yes

8.2 Offsite Power System 3 7/83 Yes ABWR and COL Applicant

Appendix A 0 7/83 Yes ABWR and COL Applicant

8.3.1 AC Power Systems (Onsite) 2 7/81 Yes

Appendix (Superseded by BTP PSB-2)

8.3.2 DC Power Systems (Onsite) 2 7/81 Yes

Appendix 8 — A Branch Technical Positions (PSB) 2 7/81 Yes

BTP ICSB 2 (PSB) (Superseded by IEEE-387)




SRP No. SRP Title

Appl.

Rev.

Issued

Date

ABWR

Appli-

cable? Comments


Rev. 0




BTP ICSB 15 (PSB) (Deleted)

BTP ICSB 17 (PSB) (Superseded by Reg. Guide 1.9)



BTP PSB 1 0 7/81 Yes

BTP PSB 2 0 7/81 Yes

Appendix 8 — B General Agenda, Station Site Visits

0 7/81 Yes

Chapter 9 Auxiliary Systems

9.1.1 New Fuel Storage 2 7/81 Yes

9.1.2 Spent Fuel Storage 3 7/81 Yes

9.1.3 Spent Fuel Pool Cooling and Cleanup System 1 7/81 Yes

9.1.4 Light Load Handling System (Related to Refueling) 2 7/81 Yes

BTP ASB 9-1 (Superseded by NUREG-0554)

9.1.5 Overhead Heavy Load Handling Systems 0 7/81 Yes

9.2.1 Station Service Water System 4 6/85 Yes ABWR and COL Applicant

9.2.2 Reactor Auxiliary Cooling Water Systems 3 6/86 Yes ABWR and COL Applicant

9.2.3 Demineralized Water Makeup System 2 7/81 Yes ABWR and COL Applicant

9.2.4 Potable and Sanitary Water Systems 2 7/81 Yes ABWR and COL Applicant

9.2.5 Ultimate Heat Sink 2 7/81 —- COL Applicant

BTP ASB 9-2 2 7/81 Yes



SRP No. SRP Title

Appl.

Rev.

Issued

Date

ABWR

Appli-

cable? Comments


Rev. 0


9.2.6 Condensate Storage Facilities 2 7/81 Yes

9.3.1 Compressed Air System 1 7/81 Yes

9.3.2 Process and Post-Accident Sampling Systems 2 7/81 Yes

9.3.3 Equipment and Floor Drainage System 2 7/81 Yes ABWR and COL Applicant

9.3.4 Chemical and Volume Control System (PWR) (Including Boron Recovery System)

2 7/81 No PWR only

9.3.5 Standby Liquid Control System (BWR) 2 7/81 Yes

9.4.1 Control Room Area Ventilation System 2 7/81 Yes ABWR and COL Applicant

9.4.2 Spent Fuel Pool Area Ventilation System 2 7/81 Yes

9.4.3 Auxiliary and Radwaste Area Ventilation System 2 7/81 Yes ABWR and COL Applicant

9.4.4 Turbine Area Ventilation System 2 7/81 Yes

9.4.5 Engineered Safety Feature Ventilation System 2 7/81 Yes

9.5.1 Fire Protection Program 3 7/81 Yes

BTP CMEB 9.5-1 2 7/81 Yes

Appendix A (Deleted)

9.5.2 Communication Systems 2 7/81 Yes ABWR and COL Applicant

9.5.3 Lighting Systems 2 7/81 Yes

9.5.4 Emergency Diesel Engine Fuel Oil Storage and Transfer System

2 7/81 Yes

9.5.5 Emergency Diesel Engine Cooling Water System 2 7/81 Yes

9.5.6 Emergency Diesel Engine Starting System 2 7/81 Yes

9.5.7 Emergency Diesel Engine Lubrication System 2 7/81 Yes

9.5.8 Emergency Diesel Engine Combustion Air Intake and Exhaust System

2 7/81 Yes



SRP No. SRP Title

Appl.

Rev.

Issued

Date

ABWR

Appli-

cable? Comments


Rev. 0


Chapter10 Steam and Power Conversion System

10.2 Turbine Generator 2 7/81 Yes

10.2.3 Turbine Disk Integrity 1 7/81 Yes

10.3 Main Steam Supply System 3 4/84 Yes

10.3.6 Steam and Feedwater System Materials 2 7/81 Yes

10.4.1 Main Condensers 2 7/81 Yes

10.4.2 Main Condenser Evacuation System 2 7/81 Yes

10.4.3 Turbine Gland Sealing System 2 7/81 Yes

10.4.4 Turbine Bypass System 2 7/81 Yes

10.4.5 Circulating Water System 2 7/81 Yes ABWR and COL Applicant

10.4.6 Condensate Cleanup System 2 7/81 Yes

10.4.7 Condensate and Feedwater System 3 4/84 Yes

BTP ASB 10-2 3 4/84 Yes

10.4.8 Steam Generator Blowdown System (PWR) 2 7/81 No PWR only

10.4.9 Auxiliary Feedwater System (PWR) 2 7/81 No PWR only

BTP ASB 10-1 2 7/81 No PWR only

Chapter 11 Radioactive Waste Management

11.1 Source Terms 2 7/81 Yes

11.2 Liquid Waste Management Systems 2 7/81 Yes

11.3 Gaseous Waste Management Systems 2 7/81 Yes

BTP ETSB 11-5 0 7/81 No

11.4 Solid Waste Management Systems 2 7/81 Yes

BTP ETSB 11-3 2 7/81 Yes

Appendix 11.4-A 0 7/81 Yes

11.5 Process and Effluent Radiological Monitoring Instrumentation and Sampling Systems

3 7/81 Yes

Appendix 11.5-A 1 7/81 Yes



SRP No. SRP Title

Appl.

Rev.

Issued

Date

ABWR

Appli-

cable? Comments


Rev. 0


Chapter 12 Radiation Protection

12.1 Assuring That Occupational Radiation Exposures Are as Low as Reasonably Achievable

2 7/81 Yes

12.2 Radiation Sources 2 7/81 Yes

12.3–12.4 Radiation Protection Design Features 2 7/81 Yes

12.5 Operational Radiation Protection Program 2 7/81 — COL Applicant

Chapter 13 Conduct of Operations

13.1.1 Management and Technical Support Organization 2 7/81 — COL Applicant

13.1.2–13.1.3

Operating Organization 2 7/81 — COL Applicant

13.2 Training (Replaced by SRP Sections 13.2.1 and 13.2.2)

13.2.1 Reactor Operator Training 0 7/81 — COL Applicant

13.2.2 Training For Non-Licensed Plant Staff 0 7/81 — COL Applicant

13.3 Emergency Planning 2 7/81 — COL Applicant

13.4 Operational Review 2 7/81 — COL Applicant

13.5 Plant Procedures (Replaced by SRP Sections 13.5.1 and 13.5.2)

13.5.1 Administration Procedures 0 7/81 — COL Applicant

13.5.2 Operating and Maintenance Procedures 1 7/85 — COL Applicant


13.6 Physical Security 2 7/81 Yes ABWR and COL Applicant



SRP No. SRP Title

Appl.

Rev.

Issued

Date

ABWR

Appli-

cable? Comments


Rev. 0


Chapter 14 Initial Test Program

14.1 Initial Plant Test Programs—PSAR (Deleted)

14.2 Initial Plant Test Programs—FSAR 2 7/81 Yes

14.3 Standard Plant Design, Initial Test Program—Final Design Approval (FDA) (Deleted)

Chapter 15 Accident Analysis

15.0 Introduction 2 7/81 Yes

15.1.1–15.1.4

Decrease in Feedwater Temperature, Increase in Feedwater Flow, Increase in Steam Flow, and Inadvertent Opening of a Steam Generator Relief or Safety Valve

1 7/81 Yes

15.1.5 Steam System Piping Failures Inside and Outside of Containment (PWR)

2 7/81 No PWR only

Appendix A 2 7/81 No PWR only

15.2.1–15.2.5

Loss of External Load, Turbine Trip, Loss of Condenser Vacuum, Closure of Main Steam Isolation Valve (BWR), and Steam Pressure Regulator Failure (Closed)

1 7/81 Yes

15.2.6 Loss of Nonemergency AC Power to the Station Auxiliaries

1 7/81 Yes

15.2.7 Loss of Normal Feedwater Flow 1 7/81 Yes

15.2.8 Feedwater system Pipe Breaks Inside and Outside Containment (PWR)

1 7/81 No PWR only

15.3.1–15.3.2

Loss of Forced Reactor Coolant Flow Including Trip of Pump and Flow Controller Malfunctions

1 7/81 Yes

15.3.3–15.3.4

Reactor Coolant Pump Rotor Seizure and Reactor Coolant Pump Shaft Break

2 7/81 Yes

15.4.1 Uncontrolled control Rod Assembly Withdrawal from a Subcritical of Low Power Startup Condition

2 7/81 Yes

15.4.2 Uncontrolled Control Rod Assembly Withdrawal at Power

2 7/81 Yes

15.4.3 Control Rod Misoperation (System Malfunction or Operator Error)

2 7/81 Yes



SRP No. SRP Title

Appl.

Rev.

Issued

Date

ABWR

Appli-

cable? Comments


Rev. 0


15.4.4–15.4.5

Startup of an Inactive Loop or Recirculation Loop at an Incorrect Temperature, and Flow Controller Malfunction Causing an Increase in BWR Core Flow Rate

1 7/81 Yes

15.4.6 Chemical and Volume Control System Malfunction That Results in a Decrease in the Boron Concentration in the Reactor Coolant (PWR)

1 7/81 No PWR only

15.4.7 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position

1 7/81 Yes

15.4.8 Spectrum of Rod Ejection Accidents (PWR) 2 7/81 No PWR only

Appendix A 1 7/81 No PWR only

15.4.9 Spectrum of Rod Drop Accidents (BWR) 2 7/81 Yes


15.5.1–15.5.2

Inadvertent Operation of ECCS and Chemical and Volume Control System Malfunction That Increases Reactor Coolant Inventory

1 7/81 Yes

15.6.1 Inadvertent Opening of a PWR Pressurizer Relief Valve or a BWR Relief Valve

1 7/81 Yes

15.6.2 Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside Containment

2 7/81 Yes

15.6.3 Radiological Consequences of Steam Generator Tube Failure (PWR)

2 7/81 No PWR only

15.6.4 Radiological Consequences of Main Steam Line Failure Outside Containment (BWR)

2 7/81 Yes

15.6.5 Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary

2 7/81 Yes



Appendix C (Deleted)

Appendix D 1 7/81 Yes

15.7.1 Waste Gas System Failure (Deleted)

15.7.2 Radioactive Liquid Waste System Leak or Failure (Released to Atmosphere) (Deleted)



SRP No. SRP Title

Appl.

Rev.

Issued

Date

ABWR

Appli-

cable? Comments


Rev. 0


Notes:

(1) See Subsection 3.8.3.2

(2) See Subsection 4.2

15.7.3 Postulated Radioactive Release Due to Liquid-Containing Tank Failures

2 7/81 Yes

15.7.4 Radiological Consequences of Fuel Handling Accidents

1 7/81 Yes

15.7.5 Spent Fuel Cask Drop Accidents 2 7/81 Yes

15.8 Anticipated Transients Without Scram 1 7/81 Yes

Appendix (Deleted)

Chapter 16 Technical Specifications

16.0 Technical Specifications 1 7/81 Yes

Chapter 17 Quality Assurance

17.1 Quality Assurance During the Design and Construction Phases

2 7/81 Yes

17.2 Quality Assurance During the Operations Phase 2 7/81 — COL Applicant

Chapter 18 Human Factors Engineering

18.0 Human Factors Engineering/Standard Review Plan Development

1 9/84 Yes

18.1 Control Room 0 9/84 Yes


18.2 Safety Parameter Display System 0 11/84 Yes




SRP No. SRP Title

Appl.

Rev.

Issued

Date

ABWR

Appli-

cable? Comments


Rev. 0


2

Table 1.8-20 NRC Regulatory Guides Applicable to ABWR

RG No. Regulatory Guide Title

Appl.

Rev.

Issued

Date

ABWR

Applicable? Comments

1.1 Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps

0 11/70 Yes

1.3 Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Boiling Water Reactors

2 6/74 Yes

1.4 Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors

2 6/74 No PWR only

1.5 Assumptions Used for Evaluating the Potential Radiological Consequences of a Steamline Break Accident for Boiling Water Reactors

0 3/71 Yes

1.6 Independence Between Redundant Standby (Onsite) Power Sources and Between Their Distribution Systems

0 3/71 Yes

1.7 Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident

2 11/78 Yes

1.8 Personnel Selection and Training -- -- -- See Table 17.0-1

1.9 Selection, Design, Qualification, and Testing of Emergency Diesel-Generator Units Used As Class 1E Onsite Electric Power Systems at Nuclear Plants

3 7/93 Yes

1.11 Instrument Lines Penetrating Primary Reactor Containment

0 3/71 Yes

1.12 Instrumentation for Earthquakes 1 4/74 No NA

1.13 Spent Fuel Storage Facility Design Basis 1 12/75 Yes

1.14 Reactor Coolant Pump Flywheel Integrity 1 8/75 No PWR only

1.16 Reporting of Operating Information —Appendix A Technical Specifications

4 8/75 --- COL Applicant

1.20 Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Testing

2 5/76 Yes


Rev. 0


1.21 Measuring, Evaluating and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light Water Nuclear Power Plants

1 6/74 Yes

1.22 Periodic Testing of Protection System Actuation-Functions

0 2/72 Yes

1.23 Onsite Meteorological Programs 0 2/72 Yes

1.24 Assumptions Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure

0 3/72 No PWR only

1.25 Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors

0 3/72 Yes

1.26 Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants

--- --- --- See Table 17.0-1

1.27 Ultimate Heat Sink for Nuclear Power Plants 2 1/76 Yes

1.28 Quality Assurance Program Requirements (Design and Construction)

--- --- --- See Table 17.0-1

1.29 Seismic Design Classification --- --- --- See Table 17.0-1

1.30 Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment

--- --- --- See Table 17.0-1

1.31 Control of Ferrite Content in Stainless Steel Weld Metal

3 4/78 Yes

1.32 Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants

2 2/77 Yes

1.33 Quality Assurance Program Requirements (Operations)


1.34 Control of Electroslag Weld Properties 0 12/72 Yes

1.35 Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containment Structures

2 1/76 Yes

Table 1.8-20 NRC Regulatory Guides Applicable to ABWR (Continued)


Appl.

Rev.

Issued

Date

ABWR



Rev. 0


1.36 Non-Metallic Insulation for Austenitic Stainless Steel

0 2/73 Yes

1.37 Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants

--- --- --- See Table 17.0-1

1.38 Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants

--- --- --- See Table 17.0-1

1.39 Housekeeping Requirements for Water-Cooled Nuclear Power Plants

--- --- --- See Table 17.0-1

1.40 Qualification Tests of Continuous-Duty Motors Installed Inside the Containment of Water-Cooled Nuclear Power Plants

0 3/73 Yes

1.41 Preoperational Testing of Redundant Onsite Electric Power Systems to Verify Proper Load Group Assignments

0 3/73 Yes

1.43 Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components

0 5/73 Yes

1.44 Control of Use of Sensitized Stainless Steel 0 5/73 Yes

1.45 Reactor Coolant Pressure Boundary Leakage Detection Systems

0 5/73 Yes

[1.47 Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems

0 5/73 Yes](4)

1.49 Power Levels of Nuclear Power Plants 1 12/73 Yes

1.50 Control of Preheat Temperature Welding of Low-Alloy Steel

0 5/73 Yes

1.52 Design, Testing, Maintenance Criteria for Post-Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants

2 3/78 Yes

1.53 Application of the Single-Failure Criterion to Nuclear Power Plant Protection Systems

0 6/73 Yes

1.54 Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants

0 6/73 Yes



Appl.

Rev.

Issued

Date

ABWR



Rev. 0


1.56 Maintenance of Water Purity in Boiling Water Reactors

1 7/78 Yes

1.57 Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components

0 6/73 Yes

1.58 Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel

Superceded See Table 17.0-1

1.59 Design Basis Floods for Nuclear Power Plants

2 8/77 Yes

1.60 Design Response Spectra for Seismic Design of Nuclear Power Plants

1 12/73 Yes

1.61 Damping Values for Seismic Design of Nuclear Power Plants

0 10/73 Yes

1.62 Manual Initiation of Protective Actions 0 10/73 Yes

1.63 Electric Penetration Assemblies in Containment Structures of Nuclear Power Plants

3 2/87 Yes

1.64 Quality Assurance Requirements for the Design of Nuclear Power Plants


1.65 Materials and Inspections for Reactor Vessel Closure Studs

0 10/73 Yes

1.68 Initial Test Programs for Water-Cooled Reactor Power Plants

2 8/78 Yes

1.68.1 Preoperational and Initial Startup Testing of Feedwater and Condensate Systems for Boiling Water Reactor Power Plants

1 1/77 Yes

1.68.2 Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants

1 7/78 Yes

1.68.3 Preoperational Testing of Instrument and Control Air Systems

0 4/82 Yes

1.69 Concrete Radiation Shields for Nuclear Power Plants

0 12/73 Yes

1.70 Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants

3 11/78 Yes

1.71 Welder Qualifications for Areas of Limited Accessibility




Appl.

Rev.

Issued

Date

ABWR



Rev. 1


1.72 Spray Pond Piping Made From Fiberglass-Reinforced Thermosetting Resin

2 11/78 Yes

1.73 Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plants

0 1/74 Yes

1.74 Quality Assurance Terms and Definitions Super-ceded

See Table 17.0-1

[1.75 Physical Independence of Electric Systems 2 9/78 Yes](4)

1.76 Design Basis Tornado for Nuclear Power Plants

0 4/74 Yes

1.77 Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors

0 5/74 No PWR only

1.78 Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release

0 6/74 Yes

1.79 Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors

1 9/75 No PWR only

1.81 Shared Emergency and Shutdown Electric Systems for Multi-Unit Power Plants

1 1/75 Yes

1.82 Water Sources for Long-Term Recirculation Cooling Following Loss-of-Coolant Accident

1 11/85 Yes

1.83 Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes

1 7/75 No PWR only

[1.84 Design and Fabrication Code Case Acceptability, ASME Section III, Division 1

27 11/90 Yes](1)

1.85 Materials Code Case Acceptability, ASME Section III, Division 1

27 11/90 Yes

1.86 Termination of Operating Licenses for Nuclear Reactors

0 6/74 ---- COL Applicant

1.87 Guidance for Construction of Class 1 Components in Elevated-Temperature Reactors (Supplement to ASME Section III Code Cases 1592, 1593, 1594, 1595, and 1596)

1 6/75 No



Appl.

Rev.

Issued

Date

ABWR



Rev. 0


1.88 Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records

Super-ceded

See Table 17.0-1

[1.89 Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants

1 6/84 Yes](2)

1.90 Inservice Inspection of Prestressed Concrete Containment Structures with Grouted Tendons


1.91 Evaluations of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants

2 2/78 Yes

[1.92 Combining Modal Responses and Spatial Components in Seismic Response Analysis

1 2/76 Yes](1)

1.93 Availability of Electric Power Sources 0 12/74 Yes

1.94 Quality Assurance Requirements for Installation, Inspection, and Testing of Structural Steel During the Construction Phase of Nuclear Power Plants

--- --- --- See Table 17.0-1

1.95 Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release

1 1/77 Yes

1.96 Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants

1 6/76 Yes

1.97 Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident

3 5/83 Yes

1.98 Assumptions for Evaluating the Potential Radiological Consequences of a Radioactive Offgas System Failure in a Boiling Water Reactor

0 3/76 Yes

1.99 Radiation Embrittlement of Reactor Vessel Materials

2 5/88 Yes

[1.100 Seismic Qualification of Electric Equipment for Nuclear Power Plants

2 6/88 Yes](2)

1.101 Emergency Planning and Preparedness for Nuclear Power Reactors

3 8/92 Yes

1.102 Flood Protection for Nuclear Power Plants 1 9/76 Yes



Appl.

Rev.

Issued

Date

ABWR



Rev. 0


[1.105 Instrument Setpoints for Safety-Related Systems

2 2/86 Yes](3)

1.106 Thermal Overload Protection for Electric Motors on Motor-Operated Valves

1 3/77 Yes

1.107 Qualifications for Cement Grouting for Prestressing Tendons in Containment Structures

1 2/77 Yes

1.108 Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants

1 8/77 Yes

1.109 Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I

1 10/77 Yes

1.110 Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Plants

0 3/76 Yes

1.111 Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors

1 7/77 Yes

1.112 Calculation for Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Power Reactors

0 5/77 Yes

1.113 Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I

1 4/77 Yes

1.114 Guidance On Being Operator At the Controls of a Nuclear Power Plant


1.115 Protection Against Low-Trajectory Turbine Missiles

1 7/77 Yes

1.116 Quality Assurance Requirements for Installation, Inspection, and Testing of Mechanical Equipment and Systems

--- --- --- See Table 17.0-1

1.117 Tornado Design Classification 1 4/78 Yes

1.118 Periodic Testing of Electric Power and Protection Systems

2 6/78 Yes

1.120 Fire Protection Guidelines for Nuclear Power Plants

1 11/87 Yes



Appl.

Rev.

Issued

Date

ABWR



Rev. 0


1.121 Bases for Plugging Degraded PWR Steam Generator Tubes

0 8/76 No PWR only

1.122 Development of floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components

1 2/78 Yes

1.123 Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants


1.124 Service Limits and Loading Combinations for Class 1 Linear-Type Component Supports

1 1/78 Yes

1.125 Physical Models for Design and Operation of Hydraulic Structures and Systems for Nuclear Power Plants

1 11/78 Yes

1.126 An Acceptable Model and Related Statistical Methods for the Analysis for Fuel Densification

1 3/78 Yes

1.127 Inspection of Water-Control Structures Associated with Nuclear Power Plants


1.128 Installation Design and Installation of Large Lead Storage Batteries for Nuclear Power Plants

1 10/78 Yes

1.129 Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Nuclear Power Plants

1 2/78 Yes

1.130 Service Limits and Loading Combination for Class 1 Plate-and-Shell-Type Component Supports

1 10/78 Yes

1.131 Qualification Tests of Electric Cable, Field Splices, and Connections for Light-Water-Cooled Nuclear Power Plants

0 8/77 Yes

1.132 Site Investigations for Foundations of Nuclear Power Plants

1 3/79 Yes

1.133 Loose-Part Detection Program for the Primary Systems of Light-Water-Cooled Reactors

1 6/81 Yes

1.134 Medical Evaluation of Licensed Personnel for Nuclear Power Plants


1.135 Normal Water Level and Discharge at Nuclear Power Plants

0 9/77 Yes



Appl.

Rev.

Issued

Date

ABWR



Rev. 0


1.136 Materials, Construction, and Testing of Concrete Containments (Articles CC-1000, -2000, and -4000 through -6000 of the “Code for Concrete Reactor Vessels and Containment”)

2 7/81 Yes

1.137 Fuel-Oil Systems for Standby Diesel Generators

1 10/79 Yes

1.138 Laboratory Investigations of Soils for Engineering Analysis and Design of Nuclear Power Plants

0 4/78 Yes

1.139 Guidance for Residual Heat Removal 0 5/78 Yes

1.140 Design, Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Absorption Units of Light-Water-Cooled Nuclear Power Plants

1 10/79 No No charcoalfiltrationrequiredfor normal operation

1.141 Containment Isolation Provisions for Fluid Systems

0 4/78 Yes

1.142 Safety-Related Concrete Structures for Nuclear Power Plants (Other Than Reactor Vessels and Containments)

1 11/81 Yes

1.143 Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants

1 10/79 Yes

1.144 Auditing of Quality Assurance Programs Nuclear Power Plants

Super-ceded

See Table 17.0-1

1.145 Atmospheric Dispersion Models for Potential Accident Consequences Assessments at Nuclear Power Plants

1 12/82 Yes

1.146 Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants

Super-ceded

See Table 17.0-1

1.147 Inservice Inspection Code Case Acceptability-ASME Section XI, Division 1

8 11/90 Yes

1.148 Functional Specifications for Active Valve Assemblies in Systems Important to Safety in Nuclear Power Plants

0 4/81 Yes

1.149 Nuclear Power Plant Simulation Facilities for Use in Operator License Examinations




Appl.

Rev.

Issued

Date

ABWR



Rev. 0


1.150 Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations

1 2/83 Yes

1.151 Instrument Sensing Lines 0 7/83 Yes

[1.152 Criteria for Programmable Digital Computer System Software in Safety-Related Systems of Nuclear Power Plants

0 11/85 Yes](4)

[1.153 Criteria for Power, Instrumentation, and Control Portions of Safety Systems

0 12/85 Yes](4)

1.154 Format and Contents of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors

0 3/87 No PWR only

1.155 Station Blackout 0 8/88 Yes

1.160 Monitoring the Effectiveness of Maintenance at Nuclear Power Plants

0 6/93 Yes

5.1 Serial Numbering of Fuel Assemblies for Light-Water-Cooled Nuclear Power Plants

0 12/72 Yes

5.7 Control of Personnel Access to Protected Areas, Vital Areas, and Material Access Areas

1 5/80 Yes

5.12 General use of Locks in the Protection and Control of Facilities and Special Nuclear Materials

0 11/73 Yes

5.44 Perimeter Intrusion Alarm Systems 2 6/80 Yes

5.65 Vital Area Access Controls, Protection of Physical Security Equipment, and Key and Lock Controls

0 9/86 Yes

8.5 Criticality and Other Interior Evacuation Signals

0 2/73 Yes

8.8 Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Reasonably Achievable

3 6/78 Yes

8.12 Criticality Accident Alarm Systems 1 2/81 Yes

8.19 Occupational Radiation Dose Assessment in Light-Water Reactor Power Plants Design Stage Man-Rem Estimates

1 6/79 Yes



Appl.

Rev.

Issued

Date

ABWR


Seismic Category / Structures 1.8-34.1

Rev. 1


Table 1.8-20 Notes:

(1) See Subsection 3.9.1.7 for restriction of change to this revision. The change restriction to R.G 1.84 applies only in regard to Code Case N-420 (See DCD/Introduction, Table 7).

(2) See Section 3.10 for restriction of change to this revision.

(3) See Subsection 7.1.2.10.9 for restriction to change this revision.

(4) See Section 7A.1(1).


Rev. 1


Table 1.8-21 Industrial Codes and Standards* Applicable to ABWR

Code or

Standard

Number Year Title

American Concrete Institute (ACI)

211.1† 1981 Practice for Selecting Proportions for Normal, Heavy Weight, and Mass Concrete.

212 1981 Guide for Admixtures in Concrete

214 1977 Recommended Practice for Evaluation of Strength Test Results of Concrete

301† 1984 Specifications for Structural Concrete for Buildings

304 1973 Practice for Measuring, Mixing, Transporting, and Placing of Concrete

305 1977 Recommended Practice for Hot Weather Concreting

306 1978 Recommended Practice for Cold Weather Concreting

307 1979 Specification for the Design and Construction of Reinforced Concrete Chimneys

308† 1981 Practice for Curing Concrete

309 1972 Practice for Consolidation of Concrete

311.1R 1981 ACI Manual of Concrete Inspection

311.4R 1981 Guide for Concrete Inspection

315† 1980 Details and Detailing of Concrete Reinforcement

318† 1989 Building Code Requirements for Reinforced Concrete

[349† 1980 Code Requirements for Nuclear Safety-Related Concrete Structures](1)

359 (See ASME BPVC Section III)

American Institute of Steel Construction (AISC)

[N690† 1984 Specifications for the Design, Fabrication, and Erection of Steel Safety-Related Structures for Nuclear Facilities](1)

-- -- Manual of Steel Construction

American Iron and Steel Insitute

SG-673 1986 Cold-Formed Steel Design Manual


Rev. 0


American Nuclear Society (ANS)

2.3† 1983 Standard for Estimating Tornado and Other Extreme Wind Characteristics at Nuclear Power Sites

2.8† 1981 Determining Design Basis Flooding at Power Reactor Sites

4.5† 1988 Criteria for Accident Monitoring Functions in Light-Water-Cooled Reactors

5.1† 1979 Decay Heat Power in LWRs

[7-4.3.2† 1982 Application Criteria for Programmable Digital Computer Systems in Safety Systems of NPGS](3)(4)

18.1 (ANSI N237) 1984 Radioactive Source Term for Normal Operation of LWRs

52.1† 1983 Nuclear Safety Design Criteria for the Design of Stationary Boiling Water Reactor Plants

55.4 1979 Gaseous Radioactive Waste Processing Systems for Light Water Reactors

56.5 1979 PWR and BWR Containment Spray System Design Criteria

56.11† 1988 Standard Design Criteria for Protection Against the Effects of Compartment Flooding in Light Water Reactor Plants

57.1†(ANSI N208) 1980 Design Requirements for LWR Fuel Handling Systems

57.2(ANSI N210) 1976 Design Requirements for LWR Spent Fuel Storage Facilities at NPP

57.3 1983 Design Requirements for New Fuel Storage Facilities at LWR Plants

[57.5† 1981 Light Water Reactor Fuel Assembly Mechanical Design and Evaluation](2)

[58.2† 1988 Design Basis for Protection of Light Water NPP Against Effects of Postulated Pipe Rupture](8)

58.8† 1984 Time Response Design Criteria for Nuclear Safety Related Operator Actions

59.51 (ANSI N195)

1976 Fuel Oil Systems for Standby Diesel-Generators

American National Standards Institute (ANSI)‡

A40 1993 National Plumbing Code

A58.1 1982 Minimum Design Loads for Buildings and other Structures, revised and redesigned as ASCE 7-1988

AG-1 (See ASME AG-1)

Table 1.8-21 Industrial Codes and Standards* Applicable to ABWR (Continued)

Code or

Standard

Number Year Title


Rev. 0


B3.5 1960 American Standard Tolerance for Ball and Roller Bearings

B30.2 (See ASME B30.2)


B30.10 (See ASME B30.10)

B30.11 (See ASME B30.11)

B30.16 (See ASME B30.16)



C1 /ASQC 1985 Specifications of General Requirements for a Quality Program

C37.01 (See IEEE C37.01)

C37.04 (See IEEE C37.04)

C37.06 1987 Preferred Ratings of Power Circuit Breakers

C37.09 (See IEEE C37.09)

C37.11 1979 Power Circuit Breaker Control Requirements

C37.13 (See IEEE C37.13)

C37.16 1988 Preferred Ratings and Related Requirements for Low Voltage AC Power Circuit Breakers

C37.17 1979 Trip Devices for AC and General-Purpose DC Low-Voltage Power Circuit Breakers

C37.20 (See IEEE C37.20)

C37.50 1989 Test Procedures for Low Voltage AC Power Circuit Breakers Used in Enclosures

C37.100 (See IEEE C37.100)

C57.12 (See IEEE C57.12)

C57.12.11 (See IEEE C57.12.11)

C57.12.80 (See IEEE C57.12.80)

C57.12.90 (See IEEE C57.12.90)

C62.41 (See IEEE C62.41)

C62.45 (See IEEE C62.45)

C63.12 (See IEEE C63.12)

D975 /ASTM 1981 Diesel Fuel Oils, Specifications for


Code or

Standard

Number Year Title


Rev. 1


HE I 1970 Standards for Steam Surface Condenser, 6th Ed., Heat Exchangers Institute

[HFS-100 1988 Human Factors Engineering of Visual Display Terminal Workstations](5)

MC11.1 1976 Quality Standard for Instrument Air

N5.12 1972 Protective Coatings (Paint) for Nuclear Industry

N13.1 1969 Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities

N14.6 1986 Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More for Nuclear Materials

N18.7 1976 Administrative Controls and Quality Assurance for the Operation Phase of Nuclear Power Plants

N45.2.1ƒ

(RG 1.37)1973 Cleaning of Fluid Systems and Associated Components During

Construction Phase of Nuclear Power Plants

N45.2.2ƒ

(RG 1.38)1972 Packaging, Shipping, Receiving, Storage, and Handling of Items

for Nuclear Power Plants During the Construction Stage

N45.2.3 1973 Housekeeping During the Construction Phase of Nuclear Power Plants

N45.2.4 1972 Quality Assurance Program Requirements for Nuclear Power Plants

N45.2.5 1974 Supplementary Quality Assurance Requirements for Installation, Inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants

N45.2.8ƒ (RG 1.116)

1976 Quality Assurance Requirements for Installation, Inspection, and Testing of Mechanical Equipment and Systems

N45.4 (See ASME N45.4)

N101.2 1972 Protective Coatings (Paints) for Light Water Nuclear Containment Facilities

N101.4 1972 QA for Protective Coatings Applied to Nuclear Facilities

N195 (See ANS 59.51)

N237 (See ANS 18.1)

N270 (See ANS 52.2)

N509 (See ASME N509)

N510 (See ASME N510)

N690 (See AISC N690)


Code or

Standard

Number Year Title


Rev. 1


NQA-1 (See ASME NQA-1)

NQA-1a (See ASME NQA-1a)

NQA-2a (See ASME NQA-2a)

OM3 1990 Requirements for preoperational and Initial Startup Vibration Test Program for Water-Cooled Power Plants

OM7 1986 Requirements for Thermal Expansion Testing of Nuclear Plant Piping Systems [September 1986 (Draft-Revision 7)]

[X3.139 1987 Fiber Distributed Data Interface (FDDI) - Token Ring Media Access Control (MAC)](3)(4)

[X3.148 1988 Fiber Distributed Data Interface (FDDI) - Token Ring Physical Layer Protocol (PHY)](3)(4)

[X3.166 1990 Fiber Distributed Data Interface (FDDI) - Physical Layer Medium Dependent (PMD)] (3)(4)

[X3T9.5/84-49 Rev. 7.1May 7, 1992

FDDI Station Mangement (SMT), Preliminary Draft](3)(4)

American Petroleum Institute (API)

620† 1986 Rules for Design and Construction of Large, Welded, Low-Pressure Storage Tanks

650† 1980 Welded Steel Tanks for Oil Storage

American Society of Heating, Refrigerating and Air-Conditioning Engineers, Inc. (ASHRAE)

30 1978 Methods of Testing Liquid Chilling Packages

33 1978 Methods of Testing Forced Circulation Air Cooling and Air Heating Coils

American Society of Mechanical Engineers (ASME)

AG-1† 1991 Code on Nuclear Air and Gas Treatment

B30.2† 1983 Overhead and Gantry Cranes (Top Running Bridge, Single or Multiple Grider, Top Running Trolley Hoist)

B30.9† 1984 Slings

B30.10† 1982 Hooks

B30.11† 1980 Monorails and Underhung Cranes

B30.16† 1981 Overhead Hoists

B31.1† 1986 Power Piping

B96.1† 1986 Specification for Welded Aluminum-Alloy Storage Tanks


Code or

Standard

Number Year Title


Rev. 1


N45.4 1972 Leakage-Rate Testing of Containment Structures for Nuclear Reactors

N509† 1989 Nuclear Power Plant Air-Cleaning Units and Components

N510† 1989 Testing of Nuclear Air-Cleaning Systems

NQA-1† 1983 Quality Assurance Program Requirements for Nuclear Facilities

NQA-1a† 1983 Addenda to ANSI/ASME NQA-1-1983

[NQA-2a† 1990 Quality Assurance Requirements of Computer Software for Nuclear Facility Application](3)(4)

OMa 1988 Operation and Maintenance of Nuclear Power Plants (Addenda to OM-1987)

Sec II 1989 BPVC Section II, Material Specifications

[Sec III 1989 BPVC Section III, Rules for Construction of Nuclear Power Plant Components](6)(8)

Sec VIII 1989 BPVC Section VIII, Rules for Construction of Pressure Vessel

Sec IX 1989 BPVC Section IX, Qualification Standard for Welding and Brazing Procedures Welder, Brazers and Welding and Brazing Operators

Sec XI 1989 BPVC Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components

American Society for Testing and Materials (ASTM)

[C776 1979 Sintered Uranium Dioxide Pellets](2)

[C934 1980 Design and Quality Assurance Practices for Nuclear Fuel Rods](2)

E84 REV. A 1991 Methods of Test of Surface Burning Characteristics of Building Materials

E119 1988 Standard Test Methods for Fire Tests of Building Construction and Materials

E152 1981 Standard Methods of Fire Tests of Door Assemblies

(See ASME BPVC Section III for ASTM Material Specifications)

American Welding Society (AWS)

A4.2† 1986 Procedures for Calibrating Magnetic Instruments to Measure the Delta Ferrite content of Anstenitic Stainless Steel Weld Metal

D1.1† 1986 Steel Structural Welding Code

D14.1† 1985 Welding of Industrial and Mill Cranes and other Material Handling Equipment


Code or

Standard

Number Year Title


Rev. 1


American Water Works Association (AWWA)

D100† 1984 Welded Steel Tanks for Water Storage

CMAA70 1983 Specification for Electric Overhead Traveling Cranes

Insulated Cable Engineer Association (ICEA)

P-46-426/IEEES-135

1982 Ampacities Including Effect of Shield Losses for Single Conductor Solid-Dielectric Power Cable 15 kV through 69 kV

P-54-440/NEMAWC-51

1987 Ampacities of Cables in Open-Top Cable Trays

S-61-402/NEMAWC-5

1973 Thermoplastic Insulated Wire & Cable for the Transmission and Distribution of Electrical Energy

S-66-524/NEMAWC-7

1982 Cross Linked Thermosetting Polyethylene Insulated Wire and Cable for Transmission and Distributor of Electrical Energy

Institute of Electrical and Electronics Engineers (IEEE)

C37.01† 1979 Application Guide for Power Circuit Breakers

C37.04† 1979 AC Power Circuit Breaker Rating Structure

C37.09† 1979 Test Procedure For Power Circuit Breakers

C37.13† 1989 Low Voltage Power Circuit Breakers

C37.20† 1987 Switchgear Assemblies and Metal-Enclosed Bus

[C37.90.2 1987 Trial-Use Standard, Withstand Capability of Relay Systems to Radiated Electromagnetic Interference form Transceivers](3)(4)

C37.100† 1992 Definitions for Power Switchgear Transformers

C57.12† 1987 General Requirements for Distribution, Power, and Regulating Transformers

C57.12.11† 1980 Guide for Installation of Oil-immersed Transformers(10MVA & Larger, 69-287 kV Rating)

C57.12.80† 1978 Terminology for Power and Distribution Transformers

C57.12.90† 1987 Test Code for Distribution, Power, and Regulating Transformers

[C62.41† 1991 Guide for Surge Voltage in Low-Voltage AC Power Circuits](3)(4)

[C62.45† 1987 Guide on Surge Testing for Equipment Connected to Low-Voltage AC Power Curcuits](3)(4)

[C63.12† 1987 American National Standard for Electromagnetic Compatibility Limits-Recommended Practice](3)(4)


Code or

Standard

Number Year Title


Rev. 1


7-4.3.2 1982 Application Criteria for Digital Computers in Safety Systems for Nuclear Facilities (to be replaced by the issued version of P 7-4.3.2, “Standard Criteria for Digital Computers Used in Safety Systems of Nuclear Power Generation Stations”)

80† 1986 Guide for Safety in AC Substation Grounding

81† 1983 Guide for Measuring Earth Resistivity, Ground Impedance, and Earth Surface Potentials of a Ground System

S-135 (See ICEA P-46-426)

141† 1986 Recommended Practice for Electric Power Distribution for Industrial Plants (IEEE Red Book)

242† 1986 Recommended Practice for Protection and Coordination of Industrial and Commercial Power Systems

[279 1971 Criteria for Protection Systems for NPGS](3)(4)

308† 1980 Criteria for Class 1E Power Systems for NPGS

317† 1983 Electrical Penetration Assemblies in Containment Structures for NPGS

[323† 1974 Qualifying Class 1E Equipment for NPGS](3)(4)(7)

334† 1974 Motors for NPGS, Type Tests of Continuous Duty Class 1E

[338† 1977 Criteria for the Periodic Testing of NPGS Safety Systems](3)(9)

[344† 1987 Recommended Practices for Seismic Qualifications of Class 1E Equipment for NPGS](7)

352† 1987 General Principles for Reliability Analysis of Nuclear Power Generating Station Protection Systems

379† 1977 Standard Application of the Single-Failure Criterion to NPGS Safety Systems

382† 1985 Qualification of Actuators for Power-Operated Valve Assemblies with Safety-Related Functions for NPP

383† 1974 Type Test of Class 1E Cables; Field Splices and Connections for NPGS

[384† 1981 Criteria for Independence of Class 1E Equipment and Circuits](3)

387† 1984 Criteria for Diesel-Generator Units Applied as Standby Power Supplies for NPGS

399† 1990 Recommended Practice for Industrial and Commercial Power Systems Analysis (IEEE Brown Book)

450† 1987 Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations


Code or

Standard

Number Year Title


Rev. 1


484† 1987 Recommended Practice for the Design and Installation of Large Lead Storage Batteries for NPGS

485† 1983 Recommended Practice for Sizing Large Lead Storage Batteries for NPGS

500 1984 Guide to the Collection and Presentation of Electronic, Sensing Component, and Mechanical Equipment Reliability Data for Nuclear Power Generating Stations.

[518 1982 Guide for the Installation of Electrical Equipment to Minimize Electrical Noise Inputs to Controllers from External Sources](3)(4)

519† 1981 IEEE Standard Recommended Practices and Requirements for Harmonic Control in Electrical Power Systems

[603† 1980 IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations](3)

622† 1987 Recommended Practice for the Design and Installation of Electric Heat Tracing Systems in Nuclear Power Generating Stations

622A† 1984 Recommended Practice for the Design and Installation of Electric Pipe Heating Control and Alarm Systems in Nuclear Power Generating Stations

[730 1984 Standard for Software Quality Assurance Plans](3)(4)

741† 1986 Standard Criteria for the Protection of Class 1E Power Systems and Equipment in Nuclear Power Generating Stations

765† 1983 Standard for Preferred Power Supply for Nuclear Power Generating Stations

[802.2† 1985 Standards for Local Area Networks: Logic Link Control](3)

[802.5† 1985 Token Ring Access Method and Physical Layer Specifications](3)

[828† 1983 Standard for Software Configuration Management Plans](3)(4)

[829† 1983 Standard for Software Test Documentation](3)(4)

[830† 1984 Standard for Software Requirements Specifications](3)(4)

[845† 1988 Guide to Evaluation of Man-Machine Performance in Nuclear Power Generating Station Control Rooms and Other Peripheries](5)

944† 1986 Recommended Practice for the Application and Testing of Uninterruptable Power Supplies for Power Generating Station

946† 1985 Recommended Practice for the Design of Safety-Related DC Auxiliary Power Systems for Nuclear Power Generating Stations

[1012† 1986 Standard for Software Verification and Validation](3)(4)

[1023† 1988 IEEE Guide to the Application of Human Factors Engineering to Systems, Equipment and Facilities of Nuclear Power Generating Stations](5)


Code or

Standard

Number Year Title


Rev. 1


[1042 1987 Guide to Software Configuration Management](3)(4)

[1050 1989 Guide for Instrumentation and Control Equipment in Generating Stations](3)(4)

[1228 (Draft) 1992 Standard for Software Safety Plans](3)(4)

Instrument Society of America (ISA)

S7.3† 1981 Quality Standard for Instrument Air

S67.02-80 1980 Nuclear-Safety-Related Instrument Sensing Line Piping and Tubing Standards for Use in Nuclear Power Plants

National Electrical Manufacturers Association (NEMA)

AB 1 1986 Molded Case Circuit Breakers

FB1 1977 Fittings and Support for Conduit and Cable Assemblies

ICS 1† 1983 General Standards for Industrial Control

ICS 2† 1988 Standards for Industrial Control Devices, Controllers and Assemblies

MG 1 1987 Motors and Generators

WC-5 (See ICEA S-61-402)

WC 7 (See ICEA S-66-524)

WC 51 (See ICEA P-54-440)

National Fire Protection Association (NFPA)

10† 1981 Portable Fire Extinguishers - Installation

10A 1973 Portable Fire Extinguishers - Maintenance and Use

11† 1988 Low Expansion Foam and Combined Agent Systems-Foam Extinguishing System

12† 1985 Carbon Dioxide Extinguishing Systems

13† 1985 Installation of Sprinklers Systems

14† 1986 Installation of Standpipe and Hose Systems

15† 1985 Standard for Water Spray Fixed Systems

16† 1991 Deluge Foam-Water Sprinkler and Foam-Water Spray Systems

16A† 1988 Recommended Practice for the Installation of Closed HeadFoam-Water Sprinkler Systems

20† 1990 Standard for the Installation of Centrifugal Fire Pumps

24† 1984 Private Service Mains and their Appurtenances


Code or

Standard

Number Year Title


Rev. 1


26† 1988 Recommended Practice for the Supervision of Valves Controlling Water Supplies for Fire Protection

37† 1984 Stationary Combustion Engines and Gas Turbines

70† 1987 National Electrical Code-Handbook 1987

72† 1990 Protective Signaling Systems

72D 1986 Proprietary Protective Signaling Systems

78† 1986 Lightning Protection Code

80† 1986 Fire Doors and Windows

80A† 1993 Protection of Buildings from Exterior Fire Exposures

90A† 1985 Installation of Air Conditioning and Ventilating Systems

91† 1983 Blower and Exhaust Systems

92A† 1988 Smoke Control Systems

101† 1985 Life Safety Code

251† 1985 Fire Test, Building Construction and Materials

252† 1984 Fire Tests, Door Assemblies

255† 1984 Building Materials, Test of Surface Burning Characteristics

321† 1987 Classification of Flammable Liquids

801† 1986 Facilities Handling Radioactive Materials

802† 1988 Nuclear Research Reactors

803† 1993 Fire Protection for Light Water Nuclear Power Plants

1961† 1979 Fire Hose

1963† 1985 Screw Threads and Gaskets for Fire Hose Connections

Steel Structures Painting Council (SSPC)

PA-1 1972 Shop, Field and Maintenance Painting

PA-2 1973 Measurements of Paint Film Thickness with Magnetic Gages

SP-1 1982 Solvent Cleaning

SP-5 1985 White Metal Blast Cleaning

SP-6 1986 Commercial Blast Cleaning

SP-10 1985 Near-White Blast Cleaning

U.S. Department of Defense (DOD)

[5000.2 1991 Defense Acquisition Management Policies and Procedures](5)


Code or

Standard

Number Year Title


Rev. 1


[AD/A223168 1990 System Engineering Management Guide](5)

[AR602-1 1983 Human Factors Engineering Program](5)

[DI-HFAC-80740 1989 Human Factors Engineering Program Plan](5)

[ESD-TR-86-278 1986 Guidelines for Designing User Interface Software](5)

[HDBK-761A 1990 Human Engineering Guidelines for Management Information Systems](5)

[HDBK-763 1991 Human Engineering Procedures Guide, Ch. 5-7 & Appendix. A&B](5)

[STD-2167A 1988 Defense System Software Development](3)(4)

[TOP 1-2-610 1990 Test Operating Procedure Part 1](5)

U.S. Military (MIL)

F-51068 Latest Edition

Filter, Particulate High-Efficiency, Fire-Resistant

[H-46855B 1979 Human Engineering Requirements for Military Systems, Equipment and Facilities](5)

[HDBK-217 Latest Edition

Reliability Prediction of Electronic Equipment](3)

[HDBK-251 Latest Edition

Reliability/Design: Thermal Applications](3)

[HDBK-759A 1981 Human Factors Engineering Design for Army Material](5)

STD-282 1956 Filter Units, Protective Clothing Gas-Mask Components and Related Products: Performance-Test Methods

[STD-461C 1987 Electromagnetic Emission and Susceptibility Requirements for the Control of Electromagnetic Interference](3)(4)

[STD-462 1967 Measurement of Electromagnetic Interference Characteristics](3)(4)

[STD-1472D 1989 Human Engineering Design Criteria for Military Systems, Equipment and Facilities](5)

[STD-1478 1991 Task Performance Analysis](5)

Others

ASCE 7 1988 Minimum Design Loads for Buildings and Other Structures

ERDA 76-21 1976 Testing of Ventilation Systems, Section 9 of Industrial Ventilation Systems

[IEC 801-2 1991 Electronic Capability for Industrial-Process Measurement and Control Equipment](3)

[IEC 880 1986 Software for Computers in the Safety Systems of Nuclear Power Stations](3)(4)


Code or

Standard

Number Year Title


Rev. 1


Notes:

(1) See Subsection 3.8.3.2 for restriction to use of these.

(2) See Subsection 4.2.

(3) See section 7A.1(1).

(4) See Section 7A.1(2).

(5) See Section 18E.1 for required use of this document.

(6) See Subsection 3.8.1.1.1 for specific restriction of change to this edition.

(7) See Section 3.10 for restriction of change to this revision.

(8) See Subsection 3.9.1.7 for specific restriction of change to this edition in application to piping design. See Table 3.2-3 for the restricted Subsections of this Code as applied to piping design only.

(9) See Subsection 7.1.1.2.

* The listing of a code or standard does not necessarily mean that it is applicable in its entirety.

† Also an ANSI code (i.e. ANSI/ASME, ANSI/ANS, ANSI/IEEE etc.).

‡ ANSI, ANSI/ANS, ANSI/ASME, and ANSI/IEEE codes are included here. Other codes that approved by ANSI and another organization are listed under the latter.

ƒ As modified by NRC accepted alternate positions to the related Regulatory Guide and identified in Table 2-1 of Reference 1 to Chapter 17.

[IEC 964 1989 Design for Control Rooms of Nuclear Power Plants, Bureau Central de la Commission Electrotechnique Internationale](5)

[ISO 7498 1984 Open Systems Interconnection-Basic Refence Model, as the Data Link Layer and Physical Layer](3)

OSHA 1910.179 1990 Overhead and Gantry Cranes

TEMA C 1978 Standards of Tubular Exchanger Manufacturers Association

UL-44 1983 Rubber-Insulated Wires and Cables

UL-489 1991 Molded-Case Circuit Breakers and Circuit Breaker Enclosures

UL-845 1988 Standard for Safety Motor Control Centers - Low Voltage Circuit Breakers

-- -- Crane Manufacturers Association of America, Specification No. 70

-- -- Aluminum Construction Manual by Aluminum Association

NCIG-01 Rev. 2 Visual Weld Acceptance Criteria for Structural Welding at Nuclear Power Plants

UBC 1991 Uniform Building Code


Code or

Standard

Number Year Title


Rev. 0


Table 1.8-22 Experience Information Applicable to ABWR

No.

Issue

Date Title Comment

Type: Generic Letters

80-06 4/25/80 Clarification of NRC Requirement for Emergency Response Facilities at Each Site

80-30 12/15/80 Periodic Updating of Final Safety Analysis Reports (FSARs) COL Applicant

80-31 12/22/80 Control of Heavy Loads

81-03 2/26/81 Implementation of NUREG-0313m, Rev. 1

81-04 2/25/81 Emergency Procedures and Training for Station Blackout Events COL Applicant

81-07 2/3/81 Control of Heavy Loads

81-10 2/18/81 Post-TMI Requirements for the Emergency Operations Facility

81-11 2/22/81 Error in NUREG-0619

81-20 4/1/81 Safety Concerns Associated with Pipe Breaks in the BWR Scram System

81-37 12/29/81 ODYN Code Reanalysis Requirements

81-38 11/10/81 Storage of Low-Level Radioactive Wastes at Power Reactor Sites

COLApplicant

82-09 4/20/82 Environmental Qualification of Safety-Related Electrical Equipment

82-21 10/6/82 Technical Specifications for Fire Protection Audits COL Applicant

82-22 10/30/82 Inconsistency Between Requirements of 10CFR73.40(d) and Standard Technical Specifications for Performing Audits of Safeguard Contingency Plans

82-27 11/15/82 Transmittal of NUREG-0763, “Guidelines for Confirmatory In-Plant Tests of Safety-Relief Valve Discharges for BWR Plants,” and NUREG-0783, “Suppression Pool Temperature Limits for BWR Containments.”

82-33 12/17/82 Supplement 1 to NUREG-0737

82-39 12/22/82 Problems with the Submittals of 10CFR73.21 Safeguards Information Licensing Review

COLApplicant

83-05 2/83 Safety Evaluation of “Emergency Procedure Guidelines,” Revision 2, NEDO-24934, June 1982

COLApplicant

83-07 2/16/83 The Nuclear Waste Policy Act of 1982 COL Applicant


Rev. 0


83-13 3/2/83 Clarification of Surveillance Requirements for HEPA Filters and Charcoal Absorber Units in Standard Technical Specifications on ESF Cleanup Systems

83-28 7/8/83 Required Actions Based on Generic Implications of Salem ATWS Events

83-33 10/19/83 NRC Positions on Certain Requirements of Appendix R to 10 CFR 50

COLApplicant

84-15 7/2/84 Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability

84-23 10/26/84 Reactor Vessel/Water Level Instrumentation in BWRs

85-01 1/9/85 Fire Protection Policy Steering Committee Report

86-10 4/24/86 Implementation of Fire Protection Requirements

87-06 3/13/87 Periodic Verification of Leak Tight Integrity of Pressure Isolation Valves

COLApplicant

87-09 6/4-87 Sections 3.0 and 4.0 of the Standard Technical Specifications (STS) on the Applicability of Limiting Conditions for Operations and Surveillance Requirements

88-01 1/25/88 NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping

88-02 1/20/88 Integrated Safety Assessment Program II (ISAP II)

88-14 8/8/88 Instrument Air Supply System Problems Affecting Safety-Related Equipment Past Related Correspondence:IE Notice 87-28, Supp. 1 NUREG-1275, Volume 2

88-15 9/12/88 Electric Power Systems — Inadequate Control Over Design ProcessPast Related Correspondence:IE Notice 88-45

88-16 10/4/88 Removal of Cycle-Specific Parameter Limits from Technical Specifications

88-18 10/20/88 Plant Record Storage on Optical DisksPast Related Correspondence:NUREG-0800 Reg. Guide 1.28, Rev. 3

COLApplicant

88-20 11/23/88 Individual Plant Examination for Severe Accident Vulnerabilities-10CFR Para. 50.54(f)

88-20 8/29/89 Generic 88-20 Supplement No. 1

Table 1.8-22 Experience Information Applicable to ABWR (Continued)

No.

Issue

Date Title Comment


Rev. 0


89-01 1/31/89 Implementation of programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program

COLApplicant

89-02 3/21/89 Actions to Improve the Detection of Counterfeit and Fraudulently Marketed ProductsPast Related Correspondence:EPRI-NP-5652, “Guideline for the Utilization of Commercial-Grade Items in Nuclear Safety-Related Applications”. Bulletins 87-02 and Supplements 1 and 2, 88-05 and Supplements 1 and 2, 88-10 IE Notices 87-66, 88-19, 88-35, 88-46 and Supplements 1 and 2, 88-48 and Supplement 1, 88-97

COLApplicant

89-04 4/3/89 Guidance on Developing Acceptable Inservice Testing Program COL Applicant

89-06 4/12/89 Task Action Plan Item I.D.2 – Safety Parameter Display System CFR 50.54(f)

1A.2.3

89-07 4/28/89 Power Reactor Safeguards Contingency Planning for Surface Vehicle Bombs

89-07Supp I

4/21/89 Power Reactor Safeguards Contingency Planning for Surface Vehicle Bombs

89-08 5/2/89 Erosion/Corrosion-Induced Pipe Wall Thinning

89-10 6/28/89 Safety-Related Motor-Operated Valve Testing and Surveillance COL Applicant

89-13 7/18/89 Service Water System Problems Affecting Safety-Related Equipment

COLApplicant

89-14 8/21/89 Line Item Improvements in Technical Specifications Removalof the 3.25 Limit on Extending Surveillance Intervals

89-15 8/21/89 Emergency Response Data System COL Applicant

89-16 9/1/89 Installation of a Hardened Wetwell Vent

89-18 9/6/89 Resolution of USI A-17, Systems Interactions Subsection 19B.2.59

89-19 9/20/89 Request for Action Related to Resolution of Unresolved Safety Issue A-47, “Safety Implication of Control Systems in LWR Nuclear Power Plants”, Pursuant to 10CFR50.54(f)

Subsection19B.2.17

89-22 10/19/89 Potential for Increased Roof Loads and Plant Area Flood Runoff Depth at Licensed Nuclear Power Plants Due to Recent Change in Probable Maximum Precipitation Criteria Developed By The National Weather Service


No.

Issue

Date Title Comment


Rev. 1


90-09 12/11/90 Alternative Requirements for Snubber Visual Inspection Intervals and Corrective Actions

91-03 03/06/91 Reporting of Safeguards Events COL Applicant

91-04 04/02/91 Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle

91-05 04/04/91 Licensee Commercial Grade Procurement and Dedication Programs

91-06 04/29/91 Resolution of Generic Issue A-30, “Adequacy of Safety-Related DC Power Supplies”, Pursuant to 10CFR50.54(f)

Subsection19B.2.52

91-10 07/08/91 Explosive Searches at Protected Area Portals COL Applicant

91-11 07/19/91 Resolution of Generic Issue 48, “LCOs for Class 1E Tie Breakers”, Pursuant to 10CFR50.54(f)

Subsection19B.2.52

91-14 09/23/91 Emergency Telecommunications

91-16 10/03/91 Licensed Operators' and Other Nuclear Facility Personnel Fitness for Duty

COLApplicant

91-17 10/17/91 Generic Safety Issue 29, “Bolting Degradation or Failure in Nuclear Power Plants”

Subsection19B.2.62

92-04 8/19/92 Resolution of the Issues Related to Reactor Vessel Level Instrumentation in BWRs Pursuant to 10CFR50.54(f)

Type: IE Bulletins

79-02 3/8/79 Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts

79-08 4/14/79 Events Relevant to BWR Identified During TMI Incident

80-01 1/11/80 ADS Valve Pneumatic Supply

80-03 2/6/80 Loss of Charcoal from Absorber Cells

80-05 3/10/80 Vacuum Condition Resulting in Damage to Chemical and Volume Control System (CVCS) Holdup Tanks

COLApplicant

80-06 3/13/80 ESF Reset Controls

80-08 4/7/80 Containment Lines Penetration Welds COL Applicant

80-10 5/6/80 Non-Radioactive System – Potential for Unmonitored Release COL Applicant

80-12 5/9/80 Decay Heat Removal System Operability COL Applicant

80-13 5/12/80 Cracking in Core Spray Spargers


No.

Issue

Date Title Comment


Rev. 0


80-15 6/18/80 Possible Loss of Emergency Notification System with Loss of Offsite Power

80-20 7/31/80 Westinghouse Type W-2 Switch Failures

80-21 11/6/80 Valve Yokes Supplied by Mole COL Applicant

80-22 9/11/80 Automatic Industries, Model 200-500-008 Sealed Source Containers

COLApplicant

80-24 11/21/80 Prevention of Damage due to H2O Leakage Inside Containment

NUREG/CR-4524

80-25 12/19/80 Operating Problems with Target Rock SRVs at BWRs

81-01 1/27/81 Surveillance of Mechanical Snubbers

81-02 4/9/81 Failure of Gate Type Valves to Close COL Applicant

81-02,Supp 1

8/19/81 Failure of Gate Type Valves to Close Against Differential Pressure

COLApplicant

81-03 4/10/81 Flow Blockage of Cooling Water to Safety System COL Applicant

82-04 12/3/82 Deficiencies in Primary Containment Electrical Penetration Assemblies

COLApplicant

83-06 7/22/83 Non-Conforming Materials Supplied by Tube-Line Corp. COL Applicant

84-01 2/3/84 Cracks in Boiling Water Reactor Mark I Containment Vent Header

84-03 8/24/84 Refueling Cavity Water Seal

85-03 11/15/85 Motor-Operated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings

COLApplicant

85-03,Supp 1

4/27/88 Motor-Operated Valve Common Mode Failure During Plant Transients Due to Improper Switch SettingsPast Related Correspondence:IE Bulletin 85-03, IE Notice 86-29, and IE Notice 87-01

COLApplicant

86-01 5/23/86 Minimum Flow Logic Problems That Could Disable RHR Pumps

86-03 10/8/86 Potential Failure of Multiple ECCS Pumps Due to Single Failure of Air-Operated Valve in Minimum Flow Recirculation Line

87-01 7/9/87 Thinning of Pipe Walls in Nuclear Power Plants

87-02 11/6/87 Fastener Testing to Determine Conformance with Applicable Material Specifications

COLApplicant


No.

Issue

Date Title Comment


Rev. 0


87-02,Supp 1

4/22/88 Fastener Testing to Determine Conformance with Applicable Material Specifications Past Related Correspondence:IE Notice 88-17

COLApplicant

87-02,Supp 2

6/10/88 Fastener Testing to Determine Conformance with Applicable Material Specifications

COLApplicant

88-04 5/5/88 Potential Safety-Related Pump LossPast Related Correspondence:IE Notice 87-59

88-07 6/15/88 Power Oscillations in Boiling Water Reactors (BWRs)Past Related Correspondence:IE Notice 88-39

88-07,Supp 1

12/30/88 Power Oscillations in Boiling Water Reactors (BWRs) Subsections 7.1.2.6.1.4 and7.1.2.1.1.2.2

90-01 03/09/90 Loss of Fill-Oil in Transmitters Manufactured by Rosemount

90-02 03/20/90 Loss of Thermal Margin Caused by Channel Box Bow

91-01 10/18/91 Reporting Loss of Criticality Safety Controls

Type: IE Information Notices

79-22 9/14/79 Qualifications of Control Systems COL Applicant

80-12 3/31/80 Instrumentation Failure Causes PORV Opening

80-21 5/16/80 Anchorage and Support of Safety-Related Electrical Equipment

80-22 5/28/80 Breakdowns in Contamination Control Programs COL Applicant

80-40 11/7/80 Excessive N2 Supply Pressure

80-42 11/24/80 Effect of Radiation on Hydraulic Snubber Fluid

81-05 3/13/81 Degraded DC Systems at Palisades COL Applicant

81-07 3/16/81 Potential Problem with Water Soluble Purge Dam Materials Used During Inert Gas Welding

COLApplicant

81-10 3/25/81 Inadvertent Containment Spray COL Applicant

81-20 7/13/81 Test Failures of Electrical Penetrations


No.

Issue

Date Title Comment


Rev. 0


81-21 7/21/81 Potential Loss of Direct Access to Ultimate Heat Sink COL Applicant

81-31 10/8/81 Failure of Safety Injection Valves COL Applicant

81-38 12/17/81 Potential Significant Equipment Failures Resulting from Contamination of Air-Operated Systems

COLApplicant

82-03 3/22/82 Environmental Tests of Electrical Terminal Block

82-10 3/3/82 Following Up Symptomatic Repairs COL Applicant

82-12 4/21/82 Surveillance of Hydraulic Snubbers

82-22 7/9/82 Failures in Turbine Exhaust Lines

82-23 7/16/82 Main Steam Isolation Valve Leakage

82-25 7/20/82 Failures of Hiller Actuators Upon Gradual Loss of Air Pressure

82-32 8/19/82 Contamination of Reactor Coolant System by Organics COL Applicant

82-40 9/22/82 Deficiencies in Primary Containment Electrical Penetration Assemblies

82-43 11/16/82 Deficiencies in LWR Air Filtration/Vent System

82-49 12/16/82 Correction for Sample Conditions for Air & Gas Monitor COL Applicant

83-03 1/28/83 Calibration of Liquid Level Instruments COL Applicant

83-07 3/7/83 Nonconformities with Materials Supplied by Tube Line Corp. COL Applicant

83-08 3/9/83 Component Failures Caused by Elevated DC Control Voltage

83-17 3/31/83 Electrical Control Logic Problem Resulting in Inoperable Auto Start of Emergency Diesel Generator

83-30 5/11/83 Misapplication of Generic EOP Guidelines COL Applicant

83-35 5/31/83 Fuel Movement with Control Rods Withdrawn at BWRs COL Applicant

83-44 7/1/83 Damage to Redundant Safety Equipment from Backflow Through the Equipment

83-46 7/11/83 Common Mode Valve Failures Degrade Surry's Recirculation Spray Subsystem

COLApplicant

83-50 8/1/83 Failure of Class 1E Circuit Breakers to Close


No.

Issue

Date Title Comment


Rev. 0


83-51 8/5/83 Diesel Generator Events

83-62 9/26/83 Failure of Toxic Gas Detectors Subsection 19B.2.40

83-64 9/29/83 Lead Shielding Attached to Safety-Related Systems COL Applicant

83-70 10/25/83 Vibration-Induced Valve Failures

83-70,Supp 1

3/4/85 Vibration-Induced Valve Failures

83-72 10/28/83 Environmental Qualification Testing Experience

83-75 11/3/83 Improper Control Rod Manipulation COL Applicant

83-80 11/23/83 Use of Specialized “Stiff” Pipe Clamps

84-09 2/13/84 Lessons Learned from NRC Inspections of Fire Protection Safe Shutdown Systems (10CFR50, App. R)

84-09,Rev. 1

3/7/84 Lessons Learned from NRC Inspections of Fire Protection Safe Shutdown Systems (10CFR50, App. R)

84-10 2/24/84 Motor-Operated Valve Torque Switches Set Below the Manufacturer’s Recommended Value

COLApplicant

84-17 3/5/84 Problems with Liquid Nitrogen Cooling Components Below the Nil Ductility Temperature

84-22 3/29/84 Deficiency in Comsip, Inc. Standard Bed Catalyst

84-23 4/5/84 Results of the NRC-Sponsored Qualification Methodology on ASCO Solenoid Valves

84-32 4/18/84 Auxiliary Feedwater Sparger and Pipe Hanger Damage

84-35 4/23/84 BWR Post-Scram Drywell Pressurization

84-38 5/17/84 Problems With Design, Maintenance, and Operation of Offsite Power Systems

84-47 6/15/84 Environmental Qualification Tests of Electrical Terminal Blocks

84-67 8/17/84 Recent Snubber Inservice Testing With High Failure Rates COL Applicant

84-69 8/29/84 Operation of Emergency Diesel Generators COL Applicant

84-69,Supp. 1

2/24/86 Operation of Emergency Diesel Generators COL Applicant

84-70 9/4/84 Reliance on Water Level Instrumentation with a Common Reference Leg

COLApplicant


No.

Issue

Date Title Comment


Rev. 0


84-70,Supp. 1

8/26/85 Reliance on Water Level Instrumentation with a Common Leg COL Applicant

84-76 10/19/84 Loss of All AC Power

84-87 12/3/84 Piping Thermal Deflection Induced by Stratified Flow

84-88 12/3/84 Standby Gas Treatment System Problems

84-93 12/17/84 Potential for Loss of Water from the Refueling Cavity

85-08 1/30/85 Industry Experience on Certain Materials Used in Safety-Related Equipment

85-13 2/21/85 Consequences of Using Soluble Dams COL Applicant

85-17 4/1/85 Possible Sticking of ASCO Solenoid Valves

85-17,Supp. 1

10/1/85 Possible Sticking of ASCO Solenoid Valves

85-24 3/26/85 Failures of Protective Coatings in Pipes and Heat Exchangers COL Applicant

85-25 4/2/85 Consideration of Thermal Conditions in the Design and Installation of Supports for Diesel Generator Exhaust Silencers

85-28 4/9/85 Partial Loss of AC Power and Diesel Generator Degradation

85-30 4/19/85 Microbiologically Induced Corrosion of Containment Service Water System

85-32 4/22/85 Recent Engine Failures of Emergency Diesel Generators

85-33 4/22/85 Undersized Nozzle-to-Shell Welded Joints in Tanks and Heat Exchangers Constructed Under the Rules of the ASME Boiler and Pressure Vessel Code

85-34 4/30/85 Heat Tracing Contributes to Corrosion Failure of Stainless Steel Piping

COLApplicant

85-35 4/30/85 Failure of Air Check Valves to Seat

85-35,Supp. 1

5/17/88 Failure of Air Check Valves to Seat

85-47 6/18/85 Potential Effect of Line-Induced Vibration on Certain Target Rock Solenoid-Operated Valves

85-51 7/10/85 Inadvertent Loss of Improper Actuation of Safety-Related Equipment

COLApplicant

85-59 7/17/85 Valve Stem Corrosion Failures

85-66 8/7/85 Discrepancies Between As-Built Construction Drawings and Equipment Installations

COLApplicant


No.

Issue

Date Title Comment


Rev. 0


85-76 9/19/85 Recent Water Hammer Events

85-77 9/20/85 Possible Loss of Emergency Notification System Due to Loss of AC Power

COLApplicant

85-81 10/17/85 Problems Resulting in Erroneously High Reading With Thermoluminscent Dosimeters

COLApplicant

85-84 10/30/85 Inadequate Inservice Testing of Main Steam Isolation Valves

85-85 10/31/85 Systems Interaction Event Resulting in Reactor System Safety/Relief Valve Opening Following a Fire-Protection Deluge System Malfunction

85-86 11/5/85 Lightning Strikes at Nuclear Power Generating Stations

85-87 11/18/85 Hazards of Inerting Atmospheres COL Applicant

85-89 11/19/85 Potential Loss of Solid-State Instrumentation Following Failure or Control Room Cooling

Subsection19B.2.40

85-90 11/19/85 Use of Sealing Compounds in an Operating Plant COL Applicant

85-91 11/27/85 Load Sequencers for Emergency Diesel Generators COL Applicant

85-92 12/2/85 Surveys of Wastes Before Disposal From Nuclear Reactor Facilities

COLApplicant

85-94 12/13/85 Potential for Loss of Minimum Flow Paths Leading to ECCS Pump Damage During a LOCA

85-96 12/23/85 Temporary Strainers Left Installed in Pump Suction Piping COL Applicant

86-01 1/6/86 Failure of Main Feedwater Check Valves Causes Loss of Feedwater System Integrity and Water-Hammer Damage

86-03 1/14/86 Potential Deficiencies in Environmental Qualification of Limitorque Motor Valve Operator Wiring

86-09 2/3/86 Failure of Check and Stop Valves Subjected to Low Flow Conditions

86-10 2/13/86 Safety Parameter Display System Malfunctions

86-29 4/25/86 Effects of Changing Valve Motor-Operator Switch SettingsPast Related Correspondence: IE Bulletin 85-03

COLApplicant

86-30 4/29/86 Design Limitations of Gaseous Effluent Monitoring System

86-39 5/20/86 Failures of RHR Pump Motors and Pump Internals


No.

Issue

Date Title Comment


Rev. 0


86-43 6/10/86 Problems with Silver Zeolite Sampling of Airborne Radioiodine

COLApplicant

86-48 6/13/86 Inadequate Testing of Boron Solution Concentration in the Standby Liquid Control System

86-50 6/18/86 Inadequate Testing to Detect Failures of Safety-Related Pneumatic Components or SystemsPast Related Correspondence:IE Notices 82-25, 85-35, 85-84, 85-94

86-51 6/18/86 Excessive Pneumatic Leakage in the Automatic Depressurization SystemPast Related Correspondence:IE Bulletins 80-01, 80-25; IE Notice 85-35; IE Inspection Report 50-458/84-18 (8/16/84)

86-53 6/26/86 Improper Installation of Heat Shrinkable Tubing COL Applicant

86-57 7/11/86 Operating Problems With Solenoid-Operated Valves at Nuclear Power Plants

86-60 7/28/86 Unanalyzed Post-LOCA Release PathsPast Related Correspondence:NUREG-0737

86-68 8/15/86 Stuck Control Rod

86-70 8/18/86 Potential Failure of All Emergency Diesel Generators

86-71 8/19/86 Recent Identified Problems With Limitorque Motor Operators Past Related Correspondence:IE Notice 86-03

86-76 8/20/86 Problems Noted in Control Room Emergency Ventilation SystemsPast Related Correspondence:Item III D.3.4 of NUREG-0737 Generic Issue 83, IE Notice 85-89

Subsection19B.2.40

86-83 9/16/86 Underground Pathways into Protected Areas, Vital Areas, Material Access Areas, and Controlled Access AreasPast Related Correspondence:NUREG-0908, ANSI 3.3

COLApplicant

86-87 10/10/86 Loss of Offsite Power Upon An Automatic Bus Transfer

86-89 10/16/86 Uncontrolled Rod Withdrawal Because of A Single Failure

86-96 11/20/86 Heat Exchanger Fouling Can Cause Inadequate Operability of Service Water SystemsPast Related Correspondence:IE Bulletin 81-03, IE Notice 81-21

COL Applicant


No.

Issue

Date Title Comment


Rev. 0


86-100 12/12/86 Loss of Offsite Power to Vital Buses at Salem 2

86-104 12/16/86 Unqualified Butt Splice Connectors Identified in Qualified Penetrations

86-106 12/16/86 Feedwater Line Break

86-106,Supp. 1

2/13/87 Feedwater Line BreakPast Related Correspondence:E Notice 82-22 EPRI Report NP-3944, 4/85

86-106,Supp. 2

3/18/87 Feedwater Line Break

86-106,Supp. 3

10/10/88 Feedwater Line Break

86-109 12/29/86 Diaphragm Failure in Scram Outlet Valve Causing Rod InsertionPast Related Correspondence:IE Notice 86-08

COLApplicant

87-06 1/30/87 Loss of Suction to Low-Pressure Service Water System Pumps Resulting From Loss of Siphon

COLApplicant

87-08 2/4/87 Degraded Motor Leads in Limitorque DC Motor OperatorsPast Related Correspondence:(Unrelated problems involving wiring installed in Limitorque motor actuators) IE Notices 83-72, 86-03 and 86-71

87-09 2/5/87 Emergency Diesel Generator Room Cooling DeficiencyPast Related Correspondence:IE Notice 86-50, 86-51 and 86-89

87-10 2/11/87 Potential for Water Hammer During Restart of Residual Heat Removal PumpsPast Related Correspondence:AEOD/E309, 4/83

87-13 2/24/87 Potential For High Radiation Fields Following Loss of Water From Fuel PoolPast Related Correspondence:IE Notice 84-93, IE Bulletin 84-03

87-14 3/23/87 Actuation of Fire Suppression System Causing Inoperability of Safety-Related Ventilation EquipmentPast Related Correspondence:IE Notice 83-41, 85-85, 86-106 Supp. 2

87-28 6/22/87 Air Systems Problems at U.S. Light Water ReactorsPast Related Correspondence:AEOD-C701


No.

Issue

Date Title Comment


Rev. 0


87-28,Sup. 1

12/28/88 Air Systems Problems at U.S. Light Water ReactorsPast Related Correspondence:AEOD-C701 NUREG-1275 Vol. 2

87-36 8/4/87 Significant Unexpected Erosion of Feedwater LinesPast Related Correspondence:IE Notice 82-22, 86-106 plus Supp. 1&2 IE Bulletin 87-01

87-43 9/8/87 Gaps in Neutron-Absorbing Material in High-Density Spent Fuel Storage RacksPast Related Correspondence:EPRI NP-4724

87-49 10/9/87 Deficiencies in Outside Containment Flooding Protection

87-50 10/9/87 Potential LOCA at High- and Low-Pressure COL Applicants from Fire Damage

87-59 11/17/87 Potential RHR Pump Loss

88-01 1/27/88 Safety Injection Pipe Failure

88-04 2/5/88 Inadequate Qualification and Documentation of Fire Barrier Penetration SealsPast Related Correspondence:10CFR50 Appendix R, Appendix A to BTP APCSB 9.5-1, NUREG-0800, ASTM E-119, BTP CMEB 9.5-1, Generic Letter 86-10

88-04,Supp. 1

8/9/88 Inadequate Qualification and Documentation of Fire Barrier Penetration Seals

88-05 2/12/88 Fire in Annunciator Control Cabinets

88-12 4/12/88 Overgreasing of Electrical Motor BearingsPast Related Correspondence:LER 387/84-036

COLApplicant

88-13 4/18/88 Water Hammer and Possible Piping Damage Caused by Misapplication of Kerotest Packless Metal Diaphragm Globe Valves

88-17 4/22/88 Summary of Responses to NRC Bulletin 87-01, “Thinning of Pipe Walls in Nuclear Power Plants”Past Related Correspondence:IE Bulletin 87-01; IE Notice 82-22, 86-106, 87-36

88-21 5/9/88 Inadvertent Criticality Events at Oskarshamn and at U.S. Nuclear Power Plants

COLApplicant

88-24 5/13/88 Failures of Air-Operated Valves Affecting Safety-Related SystemsPast Related Correspondence:IE Notice 87-28 & Supp. 1, NUREG-1275


No.

Issue

Date Title Comment


Rev. 0


88-27 5/18/88 Deficient Electrical Terminations Identified in Safety-Related Components

COLApplicant

88-35 6/3/88 Inadequate Licensee Performed Vendor AuditsPast Related Correspondence:IE Bulletin 88-05

COLApplicant

88-37 6/14/88 Flow Blockage of Cooling Water to Safety System Components Past Related Correspondence:IE Notice 81-21, 86-96; IE Bulletin 81-03

COLApplicant

88-39 6/15/88 LaSalle Unit 2 Loss of Recirculation Pumps With Power Oscillation EventPast Related Correspondence:Generic Issue B-19, Generic Letter 86-02

88-43 6/23/88 Solenoid Valve ProblemsPast Related Correspondence:IE Notices 85-17 & Supp. 1, 86-57; IE Circular 81-14

88-51 7/21/88 Failures of Main Steam Isolation Valves

88-61 8/11/88 Control Room Habitability-Recent Reviews of Operating Experience

Subsection19B.2.40

88-63 8/15/88 High Radiation Hazards from Irradiated Incore Detectors and Cables

COLApplicant

88-65 8/18/88 Inadvertent Drainings of Spent Fuel Pools

88-70 8/29/88 Check Valve Inservice Testing Program DeficienciesPast Related Correspondence:IE Notice 86-01, Generic Letter 87-06

88-72 9/2/88 Inadequacies in the Design of DC Motor-Operated Valves

88-76 9/19/88 Recent Discovery of a Phenomenon Not Previously Considered in the Design of Secondary Containment Pressure ControlPast Related Correspondence:NUREG-0800

88-77 9/22/88 Inadvertent Reactor Vessel Overfill

88-81 10/7/88 Failure of AMP Window Indent Kynar Splices and Thomas and Betts Nylon Wire Caps During Environmental Qualification Testing

88-85 10/14/88 Broken Retaining Block Studs on Anchor Darling Check Valves

88-86 10/21/88 Operating with Multiple Grounds in Direct Current Distribution Systems and Supplement 1

88-89 11/21/88 Degradation of Kapton Electrical InsulationPast Related Correspondence:IE Notices 87-08, 87-16


No.

Issue

Date Title Comment


Rev. 0


88-92 11/22/88 Potential for Spent Fuel Pool Draindown

88-95 12/8/88 Inadequate Procurement Requirements Imposed by Licensees on Vendors

COLApplicant

89-01 1/4/89 Valve Body ErosionPast Related Correspondence:IE Notice 88-17

89-04 1/17/89 Potential Problems from the Use of Space Heaters COL Applicant

89-07 1/25/89 Failures of Small-Diameter Tubing in Control Air, Fuel Oil, and Lube Oil Systems Which Render Emergency Diesel Generators Inoperable

89-08 1/26/89 Pump Damage Caused by Low-Flow Operation

89-10 1/27/89 Undetected Installation Errors in Main Steam Line Pipe Tunnel Differential Temperature Sensing Elements at Boiling Water Reactors

89-11 2/2/89 Failure of DC Motor-Operated Valves to Develop Rated Torque Because of Improper Cabling Sizing

89-14 2/16/89 Inadequate Dedication Process for Commercial Grade Components Which Could Lead to Common Mode Failure of a Safety System

89-16 2/16/89 Excessive Voltage Drop in DC SystemsPast Related Correspondence:Generic Letter 88-15

89-17 2/22/89 Contamination and Degradation of Safety-Related Battery Cells

89-20 2/24/89 Weld Failures in a Pump of Byron-Jackson Design

89-21 2/27/89 Changes in Performance Characteristics of Molded Case Circuit Breakers

89-26 3/7/89 Instrument Air Supply to Safety-Related EquipmentPast Related Correspondence:Generic Letter 88-14

89-30 3/15/89 High Temperature Environments at Nuclear Power Plants

89-36 4/4/89 Excessive Temperatures in Emergency Core Cooling System Piping Located Outside Containment

89-37 4/4/89 Proposed Amendments to 40CFR Part 61, Air Emission Standards for Radionuclides

89-39 4/5/89 List of Parties Excluded from Federal Procurement of Non-procurement Programs

COLApplicant

89-52 6/8/89 Potential Fire Damper Operational Problems


No.

Issue

Date Title Comment


Rev. 0


89-61 8/30/89 Failure of Borg-Warner Gate Valves to Close Against Differential Pressure

89-63 9/5/89 Possible Submergence of Electrical Circuits Located Above the Flood Level Because of Water Intrusion and Lack of Drainage

COLApplicant

89-64 9/7/89 Electrical Bus Bar Failures COL Applicant

89-66 9/11/89 Qualification Life of Solenoid Valves

89-68 9/25/89 Evaluation of Instrument Setpoints During Modifications COL Applicant

89-69 9/29/89 Loss of Thermal Margin Caused by Channel Box Bow COL Applicant

89-70 10/11/89 Possible Indications of Misrepresented Vendor Products COL Applicant

89-71 10/19/89 Diversion of the Residual Heat Removal Pump Seal Cooling Water Flow During Recirculation Operation Following a Loss-of-Coolant Accident

89-72 10/24/89 Failure of Licensed Senior Operators to Classify Emergency Events Properly

COLApplicant

89-73 11/1/89 Potential Overpressurization of Low Pressure Systems COL Applicant

89-76 11/21/89 Biofouling Agent: Zebra Mussel COL Applicant

89-77 11/21/89 Debris in Containment Emergency Sumps and Incorrect Screen Configurations

89-79 12/1/89 Degraded Coatings and Corrosion of Steel Containment Vessels

89-80 12/1/89 Potential for Water Hammer, Thermal Stratification, and Steam Binding in High-Pressure Coolant Injection Piping

89-81 12/6/89 Inadequate Control of Temporary Modifications to Safety-Related Systems

COLApplicant


No.

Issue

Date Title Comment


Rev. 0



No.

Issue

Date Title Comment

Type: IE Information Notices

89-83 12/11/89 Sustained Degraded Voltage on the Offsite Electrical Grid and Loss of Other Generating Stations as a Result of a Plant Trip

COLApplicant

89-87 12/19/89 Disabling of Emergency Diesel Generators by Their Neutral Ground-Fault Protection Circuitry

89-88 12/16/89 Recent NRC-Sponsored Testing of Motor-Operated Valves

90-02 01/22/90 Potential Degradation of Secondary Containment

90-05 01/29/90 Inter-System Discharge of Reactor Coolant

90-07 01/30/90 New Information Regarding Insulation Material Performance and Debris Blockage of PWR Containment Sumps

90-8 02/01/90 KR-85 Hazards From Decayed Fuel

90-13 03/05/90 Importance of Review and Analysis of Safeguards Event Logs COL Applicant

90-20 03/22/90 Personnel Injuries Resulting From Improper Operation of Radwaste Incinerators

COLApplicant

90-21 03/22/90 Potential Failure of Motor-Operated Butterfly Valves to Operate Because Valve Seat Friction was Underestimated

COLApplicant

90-22 03/23/90 Unanticipated Equipment Actuation Following Restoration of Power to Rosemount Transmitter Trip Units

COLApplicant

90-25 04/16/90 Loss of Vital AC Power With Subsequent Reactor Coolant System Heatup

COLApplicant

90-25Supp.1

03/11/90 Loss of Vital AC Power With Subsequent Reactor Coolant System Heatup

COLApplicant

90-26 04/24/90 Inadequate Flow of Essential Service Water to Room Coolers and Heat Exchangers for Engineered Safety-Feature Systems

COLApplicant

90-30 05/01/90 Ultrasonic Inspection Techniques for Dissimilar Metal Welds

90-33 05/09/90 Sources of Unexpected Occupational Radiation Exposure at Spent Fuel Storage Pools

COLApplicant

90-39 06/01/90 Recent Problems with Service Water Systems COL Applicant

90-40 06/05/90 Results of NRC-Sponsored Testing of Motor-Operated Valves COL Applicant

90-42 06/19/90 Failure of Electrical Power Equipment Due to Solar Magnetic Disturbances

90-47 07/27/90 Unplanned Radiation Exposures to Personnel Extremities Due to Improper Handling of Potentially Highly Radioactive Sources

COLApplicant

1.8 Conformance with Standard Review Plan and Applicability of Codes and Standards


Rev. 0


90-50 08/08/90 Minimization of Methane Gas in Plant Systems and Radwaste Shipping Containers

COLApplicant

90-53 08/16/90 Potential Failures of Auxiliary Steam Piping and the Possible Effects on the Operability of Vital Equipment

90-54 08/28/90 Summary of Requalification Program Deficiencies COL Applicant

90-61 09/20/90 Potential for Residual Heat Removal Pump Damage Caused by Parallel Pump Interaction

90-63 10/03/90 Management Attention to the Establishment and Maintenance of a Nuclear Criticality Safety Program

COLApplicant

90-67 10/29/90 Potential Security Equipment Weaknesses

90-68 10/30/90 Stress Corrosion Cracking of Reactor Coolant Pump Bolts

90-69 10/31/90 Adequacy of Emergency and Essential Lighting

90-70 11/06/90 Pump Explosions Involving Ammonium Nitrate

90-72 11/28/90 Testing of Parallel Disc Gate Valves in Europe

90-74 12/04/90 Information on Precursors to Severe Accidents

90-78 12/18/90 Previously Unidentified Release Path From Boiling Water Reactor Control Rod Hydraulic Units

90-81 12/24/90 Fitness For Duty COL Applicant

90-82 12/31/90 Requirements For Use of Nuclear Regulatory Commission-(NRC)-Approved Transport Packages For Shipment of Type A Quantities of Radioactive Material

COLApplicant

91-04 01/28/91 Reactor Scram Following Control Rod Withdrawal Associated With Low Power Turbine Testing

91-06 01/31/91 Lockup of Emergency Diesel Generator and Load Sequencer Control Circuits Preventing Restart of Tripped Emergency Diesel Generator

91-12 02/15/91 Potential Loss of Net Positive Suction Head (NPSH) of Standby Liquid Control System Pumps

91-13 03/04/91 Inadequate Testing of Emergency Diesel Generators (EDGs)

91-14 03/05/91 Recent Safety-Related Incidents at Large Irradiators

91-17 03/11/91 Fire Safety of Temporary Installation of Services COL Applicant

91-19 03/12/91 High-Energy Piping Failures Caused by Wall Thinning


No.

Issue

Date Title Comment


Rev. 0


91-22 03/19/91 Four Plant Outage Events Involving Loss of AC Power or Coolant Spills

91-23 03/26/91 Accident Radiation Overexposures to Personnel Due to Industrial Radiography Accessory Equipment Malfunctions

COLApplicant

91-29 04/15/91 Deficiencies Identified During Electrical Distribution System Functional Inspections

91-33 05/31/91 Reactor Safety Information for States During Exercises and Emergencies

COLApplicant

91-34 06/03/91 Potential Problems in Identifying Causes of Emergency Diesel Generator Malfunctions

91-37 06/10/91 Compressed Gas Cylinder Missile Hazards COL Applicant

91-38 06/13/91 Thermal Stratification in Feedwater System Piping

91-40 06/19/91 Contamination of Nonradioactive System and Resulting Possibility for Unmonitored, Uncontrolled Release to the Environment

COLApplicant

91-41 06/27/91 Potential Problems with the Use of Freeze Seals COL Applicant

91-42 07/27/91 Plant Outage Events Involving Poor Coordination Between Operations and Maintenance Personnel During Valve Testing and Manipulations

COLApplicant

91-46 07/18/91 Degradation of Emergency Diesel Generator Fuel Oil Delivery Systems

COLApplicant

91-47 08/06/91 Failure of Thermo-Lag Fire Barrier Material to Pass Fire Endurance Test

91-49 08/15/91 Enforcement of Safety Requirements for Radiographers COL Applicant

91-50 08/20/91 A Review of Water Hammer Events After 1985

91-51 08/20/91 Inadequate Fuse Control Programs COL Applicant

91-57 09/19/91 Operational Experience on Bus Transfers

91-58 09/20/91 Dependency of Offset Disc Butterfly Valve's Operation of Orientation With Respect to Flow

91-59 09/23/91 Problems With Access Authorization Programs COL Applicant

91-60 11/01/91 Reissuance of Information Notice 91-60: False Alarms of Alarm Ratemeters Because of Radio Frequency Interference

COLApplicant


No.

Issue

Date Title Comment


Rev. 0


91-61 09/30/91 Preliminary Results of Validation Testing of Motor-Operated Valve Diagnostic Equipment

91-63 10/03/91 Natural Gas Hazards at Fort St. Vrain Nuclear Generating Station

COL Applicant

91-64 10/09/91 Site Area Emergency Resulting From a Loss of Non-Class 1E Uninterruptable Power Supplies

91-65 10/17/91 Emergency Access to Low-Level Radioactive Waste Disposal Facilities

COLApplicant

91-66 10/18/91 (1) Erroneous Date in “Nuclear Safety Guide, TID-7016, Revision 2,” (NUREG/CR-0095, ORNL/NUREG/CSD-6 (1978) And (2) Thermal Scattering Data Limitation in the Cross-Section Sets Provided With the Keno and Scale Codes

91-68 10/28/91 Careful Planning Significantly Reduces the Potential Adverse Impacts of Loss of Offsite Power Events During Shutdown

COLApplicant

91-72 11/19/91 Issuance of a Revision to the EPA Manual of Protective Action Guides and Protective Actions for Nuclear Incidents

Type: IE Circulars

80-03 3/6/80 Protection from Toxic Gas Hazards COL Applicant

80-05 4/1/80 Emergency D/G Lube Oil COL Applicant

80-08 4/18/80 RPS Response Time

80-09 4/28/80 Problems with Plant Internal Communications Systems COL Applicant

80-10 4/29/80 Failure to Maintain Environmental Qualification of Equipment COL Applicant

80-11 5/13/80 Emergency Diesel Generator Lube Oil Cooler Failures COL Applicant

80-14 6/24/80 Radioactive Contamination of Demin Water System COL Applicant

80-18 8/22/80 10 CFR 50.59 Safety Evaluation for Changes to Radioactive Waste Treatment Systems

COLApplicant

81-03 3/2/81 Inoperable Seismic Monitoring Instrument COL Applicant

81-05 3/31/81 Self-Aligning Rod End Bushing for Pipe Supports COL Applicant


No.

Issue

Date Title Comment


Rev. 0


81-07 5/14/81 Control of Radioactivity Contaminated Material COL Applicant

81-08 5/29/81 Foundation Materials COL Applicant

81-09 7/10/81 Containment Effluent Water

81-11 7/24/81 Inadequate Decay Heat Removal COL Applicant

81-13 9/25/81 Torque Switch Electrical Bypass Circuit COL Applicant

81-14 11/5/81 Main Steam Isolation Valve Failures to Close COL Applicant

NUREG

0313Rev. 2

6/88 Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping

0371 10/78 Task Action Plans for Generic Activities Category A

0471 6/78 Generic Task Problem Description: Category B, C & D Tasks

0578 9/80 Performance Testing of BWR and PWR Relief and Safety Valves.

0588 12/79 Interim Staff Position On Environmental Qualification of Safety-Related Electrical Equipment

0619 4/80 BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking

0626 1/80 Generic Evaluation of Feedwater Transients and Small Break LOCA in GE-Designed Operating Plants and Near-Term Operating License Applications

0660 5/80 NRC Action Plan Developed as a Result of the TMI-2 Accident

0661Supp. 1

8/82 Safety Evaluation Report – Mark I Containment Long-Term Program – Resolution of Generic Technical Activity A-7

Subsection19B.2.3

0654 10/80 Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants

COLApplicant

0696 12/80 Functional Criteria for Emergency Response Facilities COL Applicant

0710Rev. 1

6/81 Licensing Requirements for Pending Applications for Construction Permits and Manufacturing License

0737Supp.1

12/82 Clarification of TMI Action Plan Requirements


No.

Issue

Date Title Comment


Rev. 0


0744Rev. 1

10/82 Resolution of the Task A-11 Reactor Vessel Materials Toughness Safety Issue

0800 7/81 Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition

0808 8/81 Mark II Containment Program Load Evaluation and Acceptance Criteria

0813 9/81 Draft Environmental Statement Related to the Operation of Calloway Plant, Unit No. 1

0933 4/93 A prioritization of Generic Safety Issues Appendix 19B

0977 3/83 NRC Fact-Finding Task Force Report on the ATWS Events at the Salem Nuclear Generating Station, Unit 1, on February 22 and 25, 1983

1150 6/89 Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, Vol. 1 & 2

1161 5/80 Recommended Revisions to USNRC-Seismic Design Criteria Subsection 19B.2.14

1174 5/89 Evaluation of Systems Interactions in Nuclear Power Plants Subsection 19B.2.59

1212 6/86 Status of Maintenance in the US Nuclear Power Industry, 1985, Vol. 1, 2

1216 8/86 Safety Evaluation PP2 Related to Operability and Reliability of Emergency Diesel Generators

1217 4/88 Evaluation of Safety Implications of Control Systems in LWR Nuclear Power Plants-Technical Findings Related to USI A-47

Subsection19B.2.17

1218 4/88 Regulatory Analysis for Proposed Resolution of USI A-47 Subsection 19B.2.17

1229 8/89 Regulatory Analysis for Resolution of USI A-17 Subsection 19B.2.59 & 19B.2.14

1233 9/89 Regulatory Analysis for USI A-40 Subsection 19B.2.14

1273 4/88 Containment Integrity Check-Technical Finds Regulatory Analysis

1296 2/88 Peer Review of High Level Nuclear Waste

1341 5/89 Regulatory Analysis for Resolution of Generic Issue 115, Enhancement


No.

Issue

Date Title Comment


Rev. 0


1353 4/89 Regulatory Analysis for the Resolution of Generic Issue 82, “Beyond Design Basis Accidents in Spent Fuel Pools”

Subsection19B.2.63

1370 9/89 Resolution of USI A-48 Subsection 19B.2.18

1275 2/91 Volume 6, Operating Experience Feedback Report Solenoid Operated Valve Problems

1339 6/90 Resolution of Generic Safety Issue 29: Bolting Degradation of Failure in Nuclear Power Plants

Subsection19B.2.62

CR-3922 1/85 Survey and Evaluation of System Interaction Events and Sources, Vol. 1, 2

Subsection19B.2.59

CR-4261 3/86 Assessment of Systems Interactions in Nuclear Power Plants Subsection 19B.2.59

CR-4262 5/85 Effects of Control System Failures on Transients, Accidents at a GE BWR, Vol. 1 and 2

CR-4387 12/85 Effects of Control System Failures on Transient and Accidents and Core-Melt Frequencies at a GE BWR

CR-4470 5/86 Survey and Evaluation of Vital Instrumentation and Control Power Supply Events

CR-5055 5/88 Atmospheric Diffusion for Control Room Habitability Assessment

Subsection19B.2.40

CR-5088 1/89 Fire Risk Scoping Study: Investigation of Nuclear Power Plant Fire Risk, Including Previously Unaddressed Issues.

CR-5230 4/89 Shutdown Decay Heat Removal Analysis: Plant Case Studies and Special Issues

CR-5347 6/89 Recommendations for Resolution of Public Comments on USI A-40

Subsection19B.2.14

CR-5458 12/89 Value-Impact Assess for Candidate Operating Procedure Upgrade Program

CR-4674 84/89 Precusors to Potential Severe Core Damage Accidents: Series


No.

Issue

Date Title Comment

COL License Information 1.9-1

Rev. 0


1.9 COL License Information

Tier 2 presents the ABWR Standard Plant design incorporating the Nuclear Island, Turbine Island and radwaste facility. Although this scope is essentially a total plant, there is a modest amount of information that must be addressed by the COL applicant. The purpose of this section is to identify the Tier 2 sections where descriptions of the COL license information are presented.

The COL license information is summarized in Table 1.9-1 in the order it is presented in Tier 2. An item number has been assigned to each entry to facilitate future identification.

1.9-2 COL License Information

Rev. 0


Table 1.9-1 Summary of ABWR Standard Plant

COL License Information

Item No. Subject Subsection

1.1 Design Process to Establish Detailed Design Documentation

1.1.11.1

1.1a Plant Design and Aging Management 1.2.3.1

1.2 P&ID Pipe Schedule 1.7.6.1

1.3 SRP Deviations 1.8.4.1

1.4 Experience Information 1.8.4.2

1.5 Emergency Procedures and Emergency Procedures Training Program

1A.3.1

1.6 Review and Modify Procedures for Removing Safety-Related Systems from Service

1A.3.2

1.7 In-plant Radiation Monitoring 1A.3.3

1.8 Reporting Failures of Reactor System Relief Valves 1A.3.4

1.9 Report on ECCS Outages 1A.3.5

1.10 Procedure for Reactor Venting 1A.3.6

1.11 Testing of SRV and Discharge Piping 1A.3.7

1.12 RCIC Bypass Start System Test 1A.3.8

1.13 Station Blackout Procedures 1C.4.1

2.1 Non-Seismic Design Parameters 2.3.1.1

2.2 Seismic Design Parameters 2.3.1.2

2.3 Site Location and Description 2.3.2.1

2.4 Exclusion Area Authority and Control 2.3.2.2

2.5 Population Distribution 2.3.2.3

2.6 Identification of Potential Hazards in Site Vicinity 2.3.2.4

2.7 Evaluation of Potential Accidents 2.3.2.5

2.8 External Impact Hazards 2.3.2.6

2.9 Local Meteorology 2.3.2.7

2.10 Onsite Meteorological Measurements Program 2.3.2.8

2.11 Short-Term Dispersion Estimates for Accident Atmosphere Releases

2.3.2.9

2.12 Long-Term Diffusion Estimates 2.3.2.10

2.13 Hydrologic Description 2.3.2.11

2.14 Floods 2.3.2.12


Rev. 0


2.15 Probable Maximum Flood on Streams and Rivers 2.3.2.13

2.16 Ice Effects 2.3.2.14

2.17 Cooling Water Channels and Reservoirs 2.3.2.15

2.18 Channel Division 2.3.2.16

2.19 Flooding Protection Requirements 2.3.2.17

2.20 Cooling Water Supply 2.3.2.18

2.21 Accidental Release of Liquid Effluents in Ground and Surface Waters

2.3.2.19

2.22 Technical Specifications and Emergency Operation Requirement

2.3.2.20

2.23 Basic Geological and Seismic Information 2.3.2.21

2.24 Vibratory Ground Motion 2.3.2.22

2.25 Surface Faulting 2.3.2.23

2.26 Stability of Subsurface Material and Foundation 2.3.2.24

2.27 Site and Facilities 2.3.2.25

2.28 Field Investigations 2.3.2.26

2.29 Laboratory Investigations 2.3.2.27

2.30 Subsurface Conditions 2.3.2.28

2.31 Evacuation and Backfilling for Foundation Construction 2.3.2.29

2.32 Effect of Groundwater 2.3.2.30

2.33 Liquefaction Potential 2.3.2.31

2.34 Response of Soil and Rock to Dynamic Loading 2.3.2.32

2.35 Minimum Static Bearing Capacity 2.3.2.33

2.36 Earth Pressures 2.3.2.34

2.37 Soil Properties for Seismic Analysis of Buried Pipes 2.3.2.35

2.38 Static and Dynamic Stability of Facilities 2.3.2.36

2.39 Subsurface Instrumentation 2.3.2.37

2.40 Stability of Slopes 2.3.2.38

2.41 Embankments and Dams 2.3.2.39

2.42 CRAC 2 Computer Code Calculations 2.3.3

3.1 Site-Specific Design Basis Wind 3.3.3.1

3.2 Site-Specific Design Basis Tornado 3.3.3.2


COL License Information (Continued)



Rev. 0


3.3 Effect of Remainder of Plant Structures, Systems and Components Not Designed for Wind Loads

3.3.3.3

3.4 Effects of Remainder of Plant Structures, Systems and Components Not Designed for Tornado Loads

3.3.3.4

3.5 Flood Elevation 3.4.3.1

3.6 Ground Water Elevation 3.4.3.2

3.7 Flood Protection Requirements for Other Structures 3.4.3.3

3.8 Not Used

3.9 Protection of Ultimate Heat Sink 3.5.4.1

3.10 Missiles Generated by Other Natural Phenomena 3.5.4.2

3.11 Site Proximity Missiles and Aircraft Hazards 3.5.4.3

3.12 Impact of Failure of Out of ABWR Standard Plant Scope Non-Safety-Related Structures, Systems, and Components Due to Design Basis Tornado

3.5.4.4

3.13 Turbine System Maintenance Program 3.5.4.5

3.14 Maintenance Equipment Missile Prevention Inside Containment

3.5.4.6

3.15 Failure of Structures, Systems, and Components Outside ABWR Standard Plant Scope

3.5.4.7

3.16 Details of Pipe Break Analysis Results and Protection Methods

3.6.5.1

3.17 Leak-Before-Break Analysis Report 3.6.5.2

3.18 Inservice Inspection of Piping in Containment Penetration Areas

3.6.5.3

3.19 Seismic Design Parameters 3.7.5.1

3.20 Pre-Earthquake Planning and Post-Earthquake Actions 3.7.5.2

3.21 Piping Analysis, Modeling of Piping Supports 3.7.5.3

3.22 Assessment of Interaction Due to Seismic Effects 3.7.5.4

3.23 Foundation Waterproofing 3.8.6.1

3.24 Site Specific Physical Properties and Foundation Settlement

3.8.6.2

3.25 Structural Integrity Pressure Results 3.8.6.3

3.26 Identification of Seismic Category I Structures 3.8.6.4

3.27 Reactor Internals Vibration Analysis, Measurement and Inspection Programs

3.9.7.1





Rev. 0


3.28 ASME Class 2 or 3 Quality Group Components with 60-Year Design Life

3.9.7.2

3.29 Pump and Valve Testing Program 3.9.7.3

3.30 Audits of Design Specifications and Design Reports 3.9.7.4

3.31 Not Used 3.9.7.5

3.32 Not Used 3.9.7.6

3.33 Not Used 3.9.7.7

3.34 Not Used 3.9.7.8

3.35 Not Used 3.9.7.9

3.36 Not Used 3.9.7.10

3.37 Equipment Qualification 3.10.5.1

3.38 Dynamic Qualification Report 3.10.5.2

3.39 Qualification by Experience 3.10.5.3

3.40 Environmental Qualification Document (EQD) 3.11.6.1

3.41 Environmental Qualification Records 3.11.6.2

3.42 Surveillance, Maintenance, and Experience Information 3.11.6.3

3.43 Radiation Environment Conditions 3I.3.3.1

4.1 Thermal Hydraulic Stability 4.3.5.1

4.2 Power/Flow Operating Map 4.4.7.1

4.3 Thermal Limits 4.4.7.2

4.4 CRD Inspection Program 4.5.3.1

4.5 CRD and FMCRD Installation and Verification During Maintenance

4.6.6.1

5.1 Conversion of Indicators 5.2.6.1

5.2 Plant Specific ISI/PSI 5.2.6.2

5.3 Reactor Vessel Water Level Instrumentation 5.2.6.3

5.4 Fracture Toughness Data 5.3.4.1

5.5 Materials and Surveillance Capsule 5.3.4.2

5.6 Plant Specific Pressure-Temperature Information 5.3.4.3

5.7 Testing of Mainsteam Isolation Valves 5.4.15.1

5.8 Analyses of 8-hour RCIC Capability 5.4.15.2

5.9 ACIWA Flow Reduction 5.4.15.3





Rev. 0


5.10 RIP Installation and Verification During Maintenance 5.4.15.4

6.1 Protection Coatings and Organic Materials 6.1.3.1

6.2 Alternate Hydrogen Control 6.2.7.1

6.3 Administrative Control Maintaining Containment Isolation

6.2.7.2

6.4 Suppression Pool Cleanliness 6.2.7.3

6.5 Wetwell-to-Drywell Vacuum Breaker Protection 6.2.7.4

6.5a Containment Penetration Leakage Test (Type B) 6.2.7.5

6.6 ECCS Performance Results 6.3.6.1

6.7 ECCS Testing Requirements 6.3.6.2

6.7a Limiting Break Results 6.3.6.3

6.8 Toxic Gases 6.4.7.1

6.9 SGTS Performance 6.5.5.1

6.9a SGTS Exceeding 90 Hours of Operation per Year 6.5.5.2

6.10 PSI and ISI Program Plans 6.6.9.1

6.11 Access Requirement 6.6.9.2

7.1 Cooling Temperature Profiles for Class 1E Digital Equipment

7.3.3.1

7.2 APRM Oscillation Monitoring Logic 7.6.3.1

7.3 Effects of Station Blackout on HVAC 7.8.1

7.4 Electrostatic Discharge on Exposed Equipment Components

7.8.2

7.5 Localized High Heat Spots in Semiconductor Material for Computing Devices

7.8.3

8.1 Diesel Generator Reliability 8.1.4.1

8.2 Periodic Testing of Offsite Equipment 8.2.4.1

8.3 Procedures When a Reserve or Unit Auxiliary Transformer is Out of Service

8.2.4.2

8.4 Offsite Power Systems Design Bases 8.2.4.3

8.5 Offsite Power Systems Scope Split 8.2.4.4

8.6 Capacity of Auxiliary Transformers 8.2.4.5

8.7 Not Used 8.3.4.1

8.8 Diesel Generator Design Details 8.3.4.2





Rev. 0


8.9 Not Used 8.3.4.3

8.10 Protective Devices for Electrical Penetration Assemblies 8.3.4.4

8.11 Not Used 8.3.4.5

8.12 Not Used 8.3.4.6

8.13 Not Used 8.3.4.7

8.14 Not Used 8.3.4.8

8.15 Offsite Power Supply Arrangements 8.3.4.9

8.16 Not Used 8.3.4.10

8.17 Not Used 8.3.4.11

8.18 Not Used 8.3.4.12

8.19 Load Testing of Class 1E Switchgear and Motor Control Centers

8.3.4.13

8.20 Administrative Controls for Bus Grounding Circuit Breakers

8.3.4.14

8.21 Administrative Controls for Manual Interconnections 8.3.4.15

8.22 Not Used 8.3.4.16

8.23 Common Industrial Standards Referenced in Purchase Specifications

8.3.4.17

8.24 Administrative Control for Switching 125Vdc Standby Charger

8.3.4.18

8.25 Control of Access to Class 1E Power Equipment 8.3.4.19

8.26 Periodic Testing of Voltage Protection Equipment 8.3.4.20

8.27 Diesel Generator Parallel Test Mode 8.3.4.21

8.28 Periodic Testing of Diesel Generator Protective Relaying 8.3.4.22

8.29 Periodic Testing of Diesel Generator Synchronizing Interlocks

8.3.4.23

8.30 Periodic Testing of Thermal Overloads and Bypass Circuitry

8.3.4.24

8.31 Periodic Inspection/Testing of Lighting System 8.3.4.25

8.32 Controls for Limiting Potential Hazards Into Cable Chases 8.3.4.26

8.33 Periodic Testing of Class 1E Equipment Protective Relaying

8.3.4.27

8.34 Periodic Testing of CVCF Power Supplies and EPAs 8.3.4.28

8.35 Periodic Testing of Class 1E Circuit Breakers 8.3.4.29





Rev. 0


8.36 Periodic Testing of Electrical Systems & Equipment 8.3.4.30

8.37 Not Used 8.3.4.31

8.38 Class 1E Battery Installation and Maintenance Requirements

8.3.4.32

8.39 Periodic Testing of Class 1E Batteries 8.3.4.33

8.40 Periodic Testing of Class 1E CVCF Power Supplies 8.3.4.34

8.41 Periodic Testing of Class 1E Battery Chargers 8.3.4.35

8.42 Periodic Testing of Class 1E Diesel Generators 8.3.4.36

9.1 New Fuel Storage Racks Criticality Analysis 9.1.6.1

9.2 Dynamic and Impact Analysis of New Fuel Storage Racks 9.1.6.2

9.3 Spent Fuel Storage Racks Criticality Analysis 9.1.6.3

9.4 Spent Fuel Rack Load Drop Analysis 9.1.6.4

9.5 New Fuel Inspection Stand Seismic Capability 9.1.6.5

9.6 Overhead Load Handling System Information 9.1.6.6

9.7 Spent Fuel Racks Structural Evaluation 9.1.6.7

9.8 Spent Fuel Racks Thermal-Hydraulic Analysis 9.1.6.8

9.9 Spent Fuel Firewater Makeup Procedures and Training 9.1.6.9

9.10 Protection of RHR System Connections to FPC System 9.1.6.10

9.11 HECW System Refrigerator Requirements 9.2.17.1

9.12 Reactor Service Water System Requirements 9.2.17.2

9.12a Not Used 9.3.12.1

9.13 Not Used 9.3.12.2

9.14 Not Used 9.3.12.3

9.15 Radioactive Drain Transfer System 9.3.12.4

9.16 Service Building HVAC System 9.4.10.1

9.17 Radwaste Building HVAC System 9.4.10.2

9.18 Contamination of DG Combustion Air Intake 9.5.13.1

9.19 Use of Communication System in Emergencies 9.5.13.2

9.20 Maintenance and Testing Procedures for Communication Equipment

9.5.13.3

9.21 Use of Portable Hand Light in Emergency 9.5.13.4

9.22 Vendor Specific Design of Diesel Generator Auxiliaries 9.5.13.5





Rev. 0


9.23 Diesel Generator Cooling Water System Design Flow and Heat Removal Requirements

9.5.13.6

9.24 Fire Rating for Penetration Seals 9.5.13.7

9.25 Diesel Generator Requirements 9.5.13.8

9.26 Applicant Fire Protection Program 9.5.13.9

9.27 HVAC Pressure Calculations 9.5.13.10

9.28 Plant Security System Criteria 9.5.13.11

9.29 Not Used 9.5.13.12

9.30 Diesel Fuel Refueling Procedures 9.5.13.13

9.31 Portable and Fixed Emergency Communication Systems 9.5.13.14

9.32 Identification of Chemicals 9.5.13.15

9.33 NUREG/CR-0660 Diesel Generator Reliability Recommendations

9.5.13.16

9.34 Sound-Powered Telephone Units 9.5.13.17

9.35 Fire-Related Administrative Controls 9.5.13.18

9.36 Periodic Testing of Combustion Turbine Generator (CTG) 9.5.13.19

9.37 Operating Procedures for Station Blackout 9.5.13.20

9.38 Quality Assurance Requirements for CTG 9.5.13.21

10.1 Low Pressure Turbine Disk Fracture Toughness 10.2.5.1

10.2 Turbine Design Overspeed 10.2.5.2

10.3 Turbine Inservice Test and Inspection 10.2.5.3

10.4 Procedures to Avoid Steam Hammer and Discharge Loads

10.3.7.1

10.5 MSIV Leakage 10.3.7.2

10.6 Radiological Analysis of the TGSS Effluents 10.4.10.1

11.1 Plant-Specific Liquid Radwaste Information 11.2.5.1

11.2 Compliance With Appendix I to 10CFR50 11.3.11.1

11.3 Plant-Specific Solid Radwaste Information 11.4.3.1

11.4 Calculation of Radiation Release Rates 11.5.6.1

11.5 Compliance with the Regulatory Shielding Design Basis 11.5.6.2

11.6 Provisions for Isokinetic Sampling 11.5.6.3

11.7 Sampling of Radioactive Iodine and Particulates 11.5.6.4





Rev. 0


11.8 Calibration Frequencies and Techniques 11.5.6.5

12.1 Regulatory Guide 8.10 12.1.4.1


12.3 Occupational Radiation Exposure 12.1.4.3


12.5 Compliance with 10CFR20 and 10CFR50 Appendix I 12.2.3.1

12.6 Airborne Radionuclide Concentration Calculation 12.3.7.1

12.7 Operational Considerations 12.7.3.2

12.8 Requirements of 10CFR70.24 12.3.7.3

12.9 Radiation Protection Program 12.5.3.1

12.10 Compliance With Para. 50.34(f)(xxvii) of 10CFR50 and NUREG-0737 Item III.D.3.3

12.5.3.2

13.1 Incorporation of Operating Experience 13.2.3.1

13.2 Emergency Plans 13.3.1.1

13.2a Review and Audit 13.4.1

13.3 Plant Operating Procedures Development Plan 13.5.3.1

13.4 Emergency Procedures Development 13.5.3.2

13.5 Implementation of the Plan 13.5.3.3

13.6 Procedures Included in Scope of Plan 13.5.3.4

13.7 Physical Security Interfaces 13.6.3

14.1 Other Testing 14.2.13.1

14.2 Test Procedures/Startup Administrative Manual 14.2.13.2

14.3 Not Used 14.2.13.3

15.1 Anticipated Operational Occurrences (AOO) 15.0.5.1

15.2 Operating Limits 15.0.5.2

15.3 Design Basis Accidents 15.0.5.3

15.4 Radiological Effects of MSIV Closure 15.2.10.1

15.5 Mislocated Fuel Bundle Accident 15.4.11.1

15.6 Misoriented Fuel Bundle Accident 15.4.11.2

15.7 Iodine Removal Credit 15.6.7.1

15.8 Not Used

15.9 Radiological Consequences of Non-Line Break Accidents 15.7.6.1





Rev. 0


16.1 COL Information Required for Plant Specific Technical Specifications

16.1.1

17.1 QA Programs For Construction And Operation 17.0.1.1

17.2 Policy and Implementation Procedures for D-RAP 17.3.13.1

17.3 D-RAP Organization 17.3.13.2

17.4 Provision for O-RAP 17.3.13.3

18.1 HSI Design Implementation Process 18.8.1

18.2 Number of Operators Needing Controls Access 18.8.2

18.3 Automation Strategies and Their Effects on Operator Reliability

18.8.3

18.4 SPDS Integration With Related Emergency Response Capabilities

18.8.4

18.5 Standard Design Features Design Validation 18.8.5

18.6 Remote Shutdown System Design Evaluation 18.8.6

18.7 Local Valve Position Indication 18.8.7

18.8 Operator Training 18.8.8

18.9 Safety System Status Monitoring 18.8.9

18.10 PGCS Malfunction 18.8.10

18.11 Local Control Stations 18.8.11

18.12 As-Built Evaluation of MCR and RSS 18.8.12

18.13 Accident Monitoring Instrumentation 18.8.13

18.14 In-Core Cooling Instrumentation 18.8.14

18.15 Performance of Critical Tasks 18.8.15

18.16 Plant Status and Post-Accident Monitoring 18.8.16

19.1 Post Accident Recovery Procedure for Unisolated CUW Line Break

19.9.1

19.2 Confirmation of CUW Operation Beyond Design Bases 19.9.2

19.3 Event Specific Procedures for Severe External Flooding 19.9.3

19.4 Confirmation of Seismic Capacities Beyond the Plant Design Bases

19.9.4

19.5 Plant Walkdowns 19.9.5

19.6 Confirmation of Loss of AC Power Event 19.9.6

19.7 Procedures and Training for Use of AC-Independent Water Addition System

19.9.7





Rev. 3


ABWR

19.8 Action to Avoid Common-Cause Failures in the Essential Multiplexing System (EMUX) and Other Common-Cause Failures

19.9.8

19.9 Action to Mitigate Station Blackout Events 19.9.9

19.10 Actions to Reduce Risk of Internal Flooding 19.9.10

19.11 Actions to Avoid Loss of Decay Heat Removal and Minimize Shutdown Risk

19.9.11

19.12 Procedures for Operation of RCIC from Outside the Control Room

19.9.12

19.13 ECCS Test and Surveillance Intervals 19.9.13

19.14 Accident Management 19.9.14

19.15 Manual Operation of MOVs 19.9.15

19.16 High Pressure Core Flooder Discharge Valve 19.9.16

19.17 Capability of Containment Isolation Valves 19.9.17

19.18 Procedures to Ensure Sample Lines and Drywell Purge Lines Remain Closed During Operation

19.9.18

19.19 Procedures for Combustion Turbine Generator to Supply Power to Condensate Pumps

19.9.19

19.19a Actions to Assure Reliability of the Supporting RCW and Service Water Systems

19.9.20

19.19b Housing of AICWA Equipment 19.9.21

19.19c Procedures to Assure SRV Operability During Station Blackout

19.9.22

19.19d Procedures for Ensuring Integrity of Freeze Seals 19.923

19.19e Procedures for Controlling Combustibles During Shutdown

19.9.24

19.19f Outage Planning and Control 19.9.25

19.19g Reactor Service Water Systems Definition 19.9.26

19.19h Capability of Vacuum Breakers 19.9.27

19.19i Capability of the Containment Atmosphere Monitoring System

19.9.28

19.19j Plant Specific Safety-Related Issues and Vendors Operating Guidance

19.9.29

19.20 Long-Term Training Upgrade 19A.3.1





Rev. 0


19.21 Long-Term Program of Upgrading of Procedures 19A.3.2

19.22 Purge System Reliability 19A.3.3

19.23 Licensing Emergency Support Facility 19A.3.4

19.24 In-Plant Radiation Monitoring 19A.3.5

19.25 Feedback of Operating, Design and Construction Experience

19A.3.6

19.26 Organization and Staffing to Oversee Design and Construction

19A.3.7

19.27 Develop More Detailed QA Criteria 19A.3.8

19.28 COL Applicant Safety Issues 19B.3.1

19.28a Testing of Isolators 19B.3.2

19.29 Seismic Capacity 19H.5.1

19.30 PRA Update 19.9.30




/14

Response to TMI Related Matters 1A-1

Rev. 0


1A Response to TMI Related Matters

1A.1 Introduction

The investigations and studies associated with the TMI accident produced several documents specifying results and recommendations, which prompted the issuances by the NRC of various bulletins, letters, and NUREGs providing guidance and requiring specific actions by the nuclear power industry. In May 1980, the issuance of NUREG-0660 (Reference 1A-1) provided a comprehensive and integrated plan and listing requirements to correct or improve the regulation and operation of nuclear facilities based on the experience from the accident at TMI and the studies and investigations of the accident. NUREG-0737(Reference 1A-2), issued in November 1980, listed items from NUREG-0660 approved by the NRC for implementation, and included additional information concerning schedules, applicability, method of implementation review, submittal dates, and clarification of technical positions. Finally, NUREG-0718 (Reference 1A-3) was issued in June 1981 to provide guidance that the NRC staff believes should be followed to account for the lessons learned from the TMI accident.

This Appendix 1A provides GE’s responses for the ABWR Standard Plant required by Section II of the NRC Standard Review Plan, those satisfying 10CFR50.34(f) are addressed in Appendix 19A. The remaining TMI issues satisfying severe accident requirements are addressed in Appendix 19B.

1A.2 NRC Positions/Responses

1A.2.1 Short-Term Accident Analysis Procedure Revision [I.C.1(3)]

NRC Position

In letters of September 13 and 27, October 10 and 30, and November 9, 1979 (References 1A-7 through 1A-11), the Office of Nuclear Reactor Regulation required licensees of operating plants, applicants for operating licenses and licensees of plants under construction to perform analyses of transients and accidents, prepare emergency procedure guidelines, upgrade emergency procedures, including procedures for operating with natural circulation conditions, and to conduct operator retraining (see also Item I.A.2.1). Emergency procedures are required to be consistent with the actions necessary to cope with the transients and accidents analyzed. Analyses of transients and accidents were to be completed in early 1980 and implementation of procedures and retraining were to be completed 3 months after emergency procedure guidelines were established; however, some difficulty in completing these requirements has been experienced. Clarification of the scope of the task and appropriate schedule revisions are being developed. In the course of review of these matters on Babcock and Wilcox (B&W)-designed plants, the staff will follow up on the bulletin and order matters relating to analysis methods and results, as listed in NUREG-0660, Appendix C

1A-2 Response to TMI Related Matters

Rev. 0


(Table C.1, Items 3, 4, 16, 18, 24, 25, 26, 27; Table C.2, Items 4, 12, 17, 18, 19, 20; and Table C.3, Items 6, 35, 37, 38, 39, 41, 47, 55, 57).

Response

In the clarification of the NUREG-0737 requirement for reanalysis of transients and accidents and inadequate core cooling and preparation of guidelines for development of emergency procedures, NUREG-0737 states:

Owners’ group or vendor submittals may be referenced as appropriate to support this reanalysis. If owners’ group or vendor submittals have already been forwarded to the staff for review, a brief description of the submittals and justification of their adequacy to support guideline development is all that is required.

GE has participated, and continues to participate, in the BWR Owners’ Group (BWROG) program to develop emergency procedure guidelines for GE BWRs. The resulting emergency procedure guidelines are generally applicable to the ABWR, as are the transient and accident analyses. Following is a brief description of the submittals to date, and a justification of their adequacy to support guideline development.

(1) Description of Submittals

(a) NEDO-24708, “Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors”, August 1979.

(b) NEDO-24708A, Revision 1,”Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors”, December 1980. This report was issued via the letter from D. B. Waters (BWR Owners’ Group) to D. G. Eisenhut (NRC) dated March 20, 1981.

(c) “BWR Emergency Procedure Guidelines (Rev. 0)”—submitted in prepublication form June 30, 1980.

(d) “BWR Emergency Procedure Guidelines (Rev. 1)”—Issued via the letter from D. B. Waters (BWR Owners’ Group) to D. G. Eisenhut (NRC) dated January 31, 1981.

(e) “BWR Emergency Procedure Guidelines (Rev. 2)”—submitted in prepublication form June 1, 1982, Letter BWROG-8219 from T. J. Dente (BWR Owners’ Group) to D. G. Eisenhut (NRC).

(f) “BWR Emergency Procedure Guidelines (Rev. 3)”, submitted in prepublication form December 22, 1982, Letter BWROG-8262 from T. J. Dente (BWR Owners’ Group) to D. G. Eisenhut (NRC).


Rev. 0


(g) NEDO-31331, “BWR Emergency Procedure Guidelines (Rev. 4)”, submitted April 23, 1987, Letter BWROG-8717, from T. A. Pickens (BWR Owners’ Group) to T. Murley (NRC).

(2) Adequacy of Submittals

The submittals described in (1) above have been discussed and reviewed extensively among the BWR Owners’ Group, the General Electric Company, and the NRC Staff.

The NRC has extensively reviewed the latest revision (Revision 4) of the Emergency Procedures Guidelines and issued a SER, “Safety Evaluation of BWR Owners’ Group Emergency Procedure Guidelines, Revision 4, NEDO-31331, March 1987”, letter from A. C. Thadani, NRC Office of Nuclear Reactor Regulation, to D. Grace, Chairman of BWR Owners’ Group, dated September 12, 1988. The SER concludes that this document is acceptable for implementation. It further states that the SER closes all the open items carried from the previous revisions of the EPG.

GE believes that, in view of these findings, no further detailed justification of the analyses or guidelines is necessary at this time. COL license information requirements pertaining to emergency procedures are discussed in Subsection 1A.3.1.

1A.2.2 Control Room Design Reviews—Guidelines and Requirements [I.D.1(1)]

NRC Position

In accordance with task Action Plan I.D.1(1), all licensees and applicants for operating licenses will be required to conduct a detailed control room design review to identify and correct design deficiencies. This detailed control room design review is expected to take about a year. Therefore, the Office of Nuclear Reactor Regulation (NRR) requires that those applicants for operating licenses who are unable to complete this review prior to issuance of a license make preliminary assessments of their control rooms to identify significant human factors and instrumentation problems and establish a schedule approved by NRC for correcting deficiencies. These applicants will be required to complete the more detailed control room reviews on the same schedule as licensees with operating plants.

Response

The design of the main control room will utilize accepted human factors engineering principles, incorporating the results of a full systems analysis similar to that described in Appendix B of NUREG-0700. A DCRDR specified in NUREG-0737 is not required by SRP Section 18.1. Details are described in Chapter 18.


Rev. 0


1A.2.3 Control Room Design—Plant Safety Parameter Display Console [I.D.2]

NRC Position

In accordance with Task Action Plan I.D.2, each applicant and licensee shall install a Safety Parameter Display System (SPDS) that will display to operating personnel a minimum set of parameters which define the safety status of the plant. This can be attained through continuous indication of direct and derived variables as necessary to assess plant safety status.

Response

The functions of the SPDS will be integrated into the overall control room design, as permitted by SRP Section 18.2. Details are found in Chapter 18.

1A.2.4 Scope of Test Program—Preoperational and Low Power Testing [I.G.1]

NRC Position

Supplement operator training by completing the special low-power test program. Tests may be observed by other shifts or repeated on other shifts to provide training to the operators.

Response

The initial test program presents an excellent opportunity for licensed operators and other plant staff members to gain valuable experience and training and, in fact, these benefits are objectives of the program (Subsection 14.2.1). The degree to which the potential benefit is realized will depend on such plant-specific factors as the organizational makeup of the startup group and overall plant staff (Subsections 14.2.2 and 13.1), as well as how the test program is conducted (Subsection 14.2.4).

The BWR Owners’ Group response to Item I.G.1 of NUREG-0737 is documented in a letter of February 4, 1981 from D.B. Waters to D.G. Eisenhut. For the most part, this issue concerns training requirements, although in the context of the initial test program. Thus, the BWROG response primarily deals with operator training issues. The exception is Appendix E of the BWROG response which describes additional tests to be conducted during the preoperational and/or startup phase.

The specific training requirements for reactor operators are discussed in Section 13.2 of the SRP, which is outside the scope of the ABWR Standard Plant (See Table 1.9-1 for COL license information requirements.). Details are found in Chapter 13.2

The additional tests specified in Appendix E of the BWROG response are contained within the initial test program described in Chapter 14. See specifically Subsections 14.2.12.1.1(3)(a), 14.2.12.1.9(3)(j), and 14.2.12.1.44(3)(a) for the relevant testing.


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1A.2.5 Reactor Coolant System Vents [II.B.1]

NRC Position

Each applicant and licensee shall install Reactor Coolant System (RCS) and reactor vessel head high point vents remotely operated from the control room. Although the purpose of the system is to vent noncondensible gases from the RCS, which may inhibit core cooling during natural circulation, the vents must not lead to an unacceptable increase in the probability of a loss-of-coolant accident (LOCA) or a challenge to containment integrity. Since these vents form a part of the reactor coolant pressure boundary, the design of the vents shall conform to the requirements of Appendix A to 10CFR50, “General Design Criteria”. The vent system shall be designed with sufficient redundancy to assure a low probability of inadvertent or irreversible actuation.

Each licensee shall provide the following information concerning the design and operation of the high point vent system:

(1) Submit a description of the design, location, size, and power supply for the vent system along with results of analyses for LOCAs initiated by a break in the vent pipe. The results of the analyses should demonstrate compliance with the acceptance criteria of 10CFR50.46.

(2) Submit procedures and supporting analyses for operator use of the vents that also include the information available to the operator for initiating or terminating vent usage.

Response

The capability to vent the ABWR reactor coolant system is provided by the safety/relief valves (SRVs) and reactor coolant vent line, as well as other systems. The COL applicant will develop plant-specific procedures to govern the operator’s use of the relief mode for venting the reactor (Subsection 1A.3.6). The capability of these systems and their satisfaction of Item II.B.1 are discussed below.

The ABWR design includes various means of high-point venting. Among these are:

(1) Normally closed reactor vessel head vent valves, operable from the control room, which discharge to the drywell. The reactor coolant vent line is located at the very top of the reactor vessel as shown in the nuclear boiler system P&ID (Figure 5.1-3). This 50A line contains two safety-related Class 1E motor-operated valves that are operated from the control room. The location of this line permits it to vent the entire reactor core system normally connected to the reactor pressure vessel. In addition, since this vent line is part of the original design, it has already been considered in all the design basis accident (DBA) analyses contained elsewhere in this document.

(2) Normally open reactor head vent line, which discharges to a main steamline.


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The conclusions from this vent evaluation are as follows:

(1) Reactor vessel head vent valves exist to relieve head pressure (at shutdown) to the drywell via remote operator action.

(2) The reactor vessel head is continuously swept to the main condenser and can be vented during operating conditions.

(3) The size of the vents is not a critical issue because BWR SRVs have substantial capacity, exceeding the full power steaming rate of the nuclear boiler.

(4) No new 10CFR50.46 conformance calculations are required, because the vent provisions are part of the plant’s original design and are covered by the original design bases.

(5) Plant-specific procedures govern the operator’s use of the relief mode for venting reactor pressure.

1A.2.6 Plant Shielding to Provide Access to Vital Areas and Protect Safety Equipment forPost-Accident Operation [II.B.2]

NRC Position

With the assumption of a post-accident release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalent of 50% of the core radioiodine, 100% of the core noble gas inventory, and 1% of the core solids are contained in the primary coolant), each licensee shall perform a radiation and shielding-design review of the spaces around systems that may, as a result of an accident, contain highly radioactive materials. The design review should identify the location of vital areas and equipment, such as the control room, radwaste control stations, emergency power supplies, motor control centers, and instrument areas, in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during post-accident operation of these systems.

Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or post-accident procedural controls. The design review shall determine which types of corrective actions are needed for vital areas throughout the facility.

Response

A review of the radiation and shielding of the ABWR Standard Plant post- accident operations has been made. It has been found that there is adequate access to vital areas and that safety equipment is adequately protected. No need for corrective action was identified. Details of the review may be found in Attachment A to Appendix 1A.


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1A.2.7 Post-Accident Sampling [II.B.3]

NRC Position

A design and operational review of the reactor coolant and containment atmosphere sampling line systems shall be performed to determine the capability of personnel to obtain (less than 1 hour) a sample under accident conditions without incurring a radiation exposure to any individual in excess of 0.05 and 0.50 Sv to the whole body or extremities, respectively. Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products. If the review indicates that personnel could not promptly and safely obtain the samples, additional design features or shielding should be provided to meet the criteria.

A design and operational review of the radiological spectrum analysis facilities shall be performed to determine the capability to quantify (in less than 2 hours) certain radionuclides that are indicators of the degree of core damage. Such radionuclides are noble gases (which indicate cladding failure), iodines and cesiums (which indicate high fuel temperatures), and nonvolatile isotopes (which indicate fuel melting). The initial reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4 release. The review should also consider the effects of direct radiation from piping and components in the auxiliary building and possible contamination and direct radiation from airborne effluents. If the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria.

In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions. Procedures shall be provided to perform boron and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source term). Both analyses shall be capable of being completed promptly (i.e., the boron sample analysis within an hour and the chloride sample analysis within a shift).

Response

Discharges From Plant and Containment— During the development of an accident, samples of liquid and gaseous discharges from both the plant and containment will be obtained. Chemical and radiochemical analyses will be performed for protection of the health and safety of the public and the plant operators. These samples will be obtained from the Process Sampling System. The Post Accident Sampling Systems will not be required to obtain these samples.

Core Damage Assessment—During this initial period, instrumentation will provide sufficient information for the operators to perform their duties. For example, the containment high range radiation meters will give instant information concerning the radiation level in containment (To obtain data from the PASS several hours may be required for sampling and analyses.). Calculations can be performed to relate


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containment radiation level with the probable extent of core damage. Core damage assessment instrumentation is described in Section 18.4.6. This section describes the safety parameter display system (SPDS). The principle purpose of the SPDS is to aid the control room personnel during abnormal and emergency conditions in determining the safety status of the plant and in assessing whether abnormal conditions warrant corrective action by operators to avoid a degraded core. The following critical safety functions are provided at the wide screen display panel in the main control room:

(1) Reactivity control

(2) Reactor core cooling and heat removal from the primary system

(3) Reactor coolant system integrity

(4) Radioactivity control

(5) Contamination conditions

This instrumentation and the PASS work together to obtain complementary information. After this initial period during the development of an accident, the ABWR PASS will be used to obtain samples of reactor water and containment atmosphere to assess the extent of core damage. The ABWR PASS has been designed to safely obtain samples with radioactivity levels up to 37,000 M Bq/g. Approximately one day after a serious core damage accident, it is expected that sample radioactivity levels will be no more than this value. Early in such an accident, the plant instrumentation in the main control room would be indicating that abnormal conditions exist. If a reactor coolant sample were obtained which had excessive radioactivity, as measured by the area radiation monitor in the PASS area, the plant operators would use this high radiation information as confirmatory evidence that severe core damage has occurred and continue following the emergency operating procedures. It would not be necessary to perform any radiochemical analyses to reach this conclusion. During less severe accidents, in which only some cladding damage has occurred, samples may be obtained from either the Process Sampling System or PASS.

NUREG-0737 Requirements— The ABWR PASS has been designed to meet the eleven requirements listed in NUREG-0737 except as noted below:

(1) The combined time allotted for sampling and analysis should be 3 hours or less from the time a decision is made to take a sample. Meets the requirements of NUREG-0737.


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(2) There shall be onsite capability to perform the following within the 3 hour time period:

(a) Determine the presence and amount of certain radionuclides in the reactor coolant and containment atmosphere that may be indicators of the degree of core damage. Meets the requirements of NUREG-0737.

(b) Hydrogen in containment atmosphere. Hydrogen in containment atmosphere is measured by the Containment Atmospheric Monitoring System. Meets the requirements of NUREG-0737.

(c) Dissolved gases, chloride and boron in liquids. Dissolved gases are discussed in item 4 below. Meets the requirements concerning chloride and boron of NUREG-0737.

(d) Inline monitoring capability is acceptable. No inline monitors are provided in PASS.

(3) Sampling need not depend upon an isolated auxiliary system being put into operation. Meets the requirements of NUREG-0737.

(4) Reactor coolant samples and analyses for total dissolved gases and hydrogen are required. During a severe core damage accident for the ABWR, the reactor water will become mixed with the suppression pool water. The pressure in the reactor vessel will decrease to approximately the pressure within the wetwell and the drywell. As a result of this decrease in pressure, dissolved gases will partially pass out of the water phase into the gas phase. Data on gases dissolved in the reactor water under these conditions will have little meaning in responding to the accident. During accidents in which only a small amount of cladding damage has occurred or in accidents in which the reactor vessel has not been depressurized, pressurized reactor water samples may be obtained from the Process Sampling System. Therefore, the ability to obtain pressurized or depressurized reactor water samples for dissolved gas analyses has not been provided.

(5) If both of the following are present:

(a) There is only a single barrier between primary containment and the cooling water.

(b) If the cooling water is sea water or brackish water, chloride analysis within 24 hours after sampling shall be provided. If both are not present, the time to complete the analyses is increased to 4 days. Analysis does not have to be done onsite. Meets the requirements of NUREG-0737. (Note that there are several barriers to prevent chloride intrusion from the power cycle cooling water into the reactor vessel. These barriers are: the


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main condenser tubing, the condensate polishing demineralizers and the pumps and valves in the condensate/feedwater systems. These pumps are stopped and these valves closed during upset conditions. Thus, because both factors are not present, the time to complete the analysis is increased to 4 days.)

(6) It must be possible to obtain and analyze a sample without radiation exposures to any individual exceeding 0.05 Sv for whole body and 0.50 Sv for extremities. Meets the requirements of 50.34(f)(2)(viii).

(7) Ability to sample and analyze for reactor coolant boron must be provided. Meets the requirements of NUREG-0737.

(8) If inline monitoring is used, backup sampling and analysis capability must be provided. Inline monitoring is not used. Meets the requirements of NUREG-0737.

(9)

(a) Capability to identify and quantify a specified number of isotopes over a range of nuclide concentrations from approximately 37,000 Bq/g to 370,000 M Bq/g. Capability is provided to identify and quantify the desired isotopes in samples over a range from approximately 37,000 Bq/g to 37,000 M Bq/g. Samples obtained during the accident recovery phase would be within this range for most core damage accidents. If the gross radioactivity levels are higher than 37,000 M Bq/g, this would confirm that severe core damage has occurred.

(b) Restrict background levels of radiation in the laboratory and provide proper ventilation. Meets the requirements of NUREG-0737.

(10) Provide adequate accuracy, range and sensitivity to provide pertinent information. Meets the requirements of NUREG-0737.

(11)

(a) Provide sample lines with proper features for sampling during accident conditions. Meets the requirements of NUREG-0737.

(b) PASS ventilation exhaust should be filtered with charcoal adsorbers and HEPA filters. Meets the requirements of NUREG-0737.


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Summary

The post-accident sampling system meets the requirements of the NRC position with the following exceptions:

(1) The upper limit of activity in the samples at the time they are taken is as follows:

(a) Liquid sample 37,000 M Bq/ml

(b) Gas sample 3700 M Bq/ml.

(2) Radiological measurements could be performed 24 hours following the accident.

(3) Boron concentration measurements could be performed 8 hours following the accident.

(4) There is no need to perform chloride measurements.

(5) There is no need to analyze dissolved gases.

1A.2.8 Rule Making Proceeding or Degraded Core Accidents [II.B.8]

Response to this TMI action plan item is addressed in Appendix 19A.

1A.2.9 Coolant System Valves—Testing Requirements [II.D.1]

NRC Position

Pressurized water reactor and boiling water reactor licensees and applicants shall conduct testing to qualify the reactor coolant system relief and safety valves under expected operating conditions for design basis transients and accidents.

Response

The ABWR safety/relief valve (SRV) is postulated to discharge steam only, not liquid or two-phase flow under expected operating conditions for design basis transients and accidents.

A generic test program was conducted through the BWR Owners’ Group (Reference 1A-10) to satisfy the discharge of steam. These steam discharge test results will be used as the qualification basis for plant-specific SRV models and discharge piping that are sufficiently similar to those reported in Reference 1A-11. [Plant-specific SRV models and discharge piping that are not similar will be tested in accordance with NUREG-0737 requirements.]* See Subsection 1A.3.7 for COL license information.

The ABWR system logic for response to high water level conditions is described in Subsection 7.3.1.1.1.1(3) and is considered to be sufficiently redundant that the

* See Subsection 3.9.1.7.


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probability of steamline flooding by ECCS is extremely low. There is no high drywell pressure signal that would inhibit this logic system.

In the ABWR design, each of three RHR shutdown cooling lines has its own separate containment penetration and its own separate source of suction from the reactor vessel. Alternate shutdown using the SRV is therefore not required for the ABWR in order to meet single failure rules. Hence, the ABWR does not require SRV testing with liquid under low pressure conditions associated with this event as required in past BWRs.

1A.2.10 Relief and Safety Valve Position Indication [II.D.3]

NRC Position

Reactor Coolant System relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve-position detection device or a reliable indication of flow in the discharge pipe.

Response

The ABWR Standard Plant SRVs are equipped with position sensors which are qualified as Class 1E components. These are used to monitor valve position.

In addition, the downstream pipe from each valve line is equipped with temperature elements which signal the annunciator and the plant process computer when the temperature in the tailpipe exceeds the predetermined setpoint.

These sensors are shown on Figure 5.1-3 (Nuclear Boiler System P&ID).

1A.2.11 Systems Reliability [II.E.3.2]

This TMI action plan item superseded by USI A-45. USI A-45 is addressed in Appendix 19B.

1A.2.12 Coordinated Study of Shutdown Heat Removal Requirements [II.E.3.3]

This TMI action plan item superseded by USI A-45. USI A-45 is addressed in Appendix 19B.

1A.2.13 Containment Design—Dedicated Penetration [II.E.4.1]

NRC Position

For plant designs with external hydrogen recombiners, provide redundant dedicated containment penetrations so that, assuming a single failure, the recombiner systems can be connected to the containment atmosphere.

Response

A Flammability Control System is provided to control the concentration of oxygen in the primary containment. The FCS utilizes two permanently installed recombiners


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located in the secondary containment. The FCS is operable in the event of a single active failure. The FCS is described in Subsection 6.2.5.

1A.2.14 Containment Design—Isolation Dependability [II.E.4.2]

NRC Position

(1) Containment isolation system designs shall comply with the recommendations of the Standard Review Plan, Subsection 6.2.4 (i.e., that there be diversity in the parameters sensed for the initiation of containment isolation).

(2) All plant personnel shall give careful consideration to the definition of essential and non-essential systems, identify each system determined to be non-essential, describe the basis for selection of each essential system, modify their containment isolation designs accordingly, and report the results of the reevaluation to the NRC.

(3) All nonessential systems shall be automatically isolated by the containment isolation signal.

(4) The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves. Reopening of containment isolation valves shall require deliberate operator action.

(5) The containment setpoint pressure that initiates containment isolation for non-essential penetrations must be reduced to the minimum compatible with normal operating conditions.

(6) Containment purge valves that do not satisfy the operability criteria set forth in Branch Technical Position CSB 6-4 or the Staff Interim Position of October 23, 1979 must be sealed closed as defined in SRP 6.2.4, Item II.6.f during operational conditions 1, 2, 3, and 4. Furthermore, these valves must be verified to be closed at least every 31 days.

(7) Containment purge and vent isolation valves must close on a high radiation signal.

Response

(1) The isolation provisions described in the Standard Review Plan, Subsection 6.2.4 (i.e., that there be diversity in the parameters sensed for the initiation of containment isolation) were reviewed in conjunction with the ABWR Standard Plant design. It was determined that the ABWR Standard Plan is designed in accordance with these recommendations of the SRP.


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(2) This request appears to be directed primarily toward operating plants. However, the classification of structures, systems and components for the ABWR Standard Plant design is addressed in Section 3.2. The basis for classification is also presented in Section 3.2. The ESF system, with remote manual valves with leakage detection outside the containment, is delineated in Tables 6.2-7. The ABWR Standard Plant fully conforms with the NRC position so far as it relates to the new equipment supplier.

(3) All non-essential systems comply with the NRC position to automatically isolate by the containment isolation signals, and by redundant safety grade isolation valves.

(4) Control systems for automatic containment isolation valves are designed in accordance with this position for the ABWR Standard Plant Design.

(5) The ABWR Standard Plant design is consistent with this position.

(6) All ABWR containment purge valves meet the criteria provided in BTP CSB 6-4. The main 550A purge valves are fail-closed and are maintained closed through power operation as defined in the plant technical specifications. All purge and vent valves are remote pneumatically-operated, fail closed and receive containment isolation signals. Certain vent valves can be opened manually in the presence of an isolation signal, to permit venting through the SGTS.

(7) In the ABWR design, the containment purge and vent isolation valves will be automatically isolated on high radiation levels detected in the Reactor Building HVAC air exhaust or in the fuel handling area air exhaust.

1A.2.15 Additional Accident—Monitoring Instrumentation [II.F.1]

NRC Position

Noble gas effluent monitors shall be installed with an extended range designed to function during accident conditions, as well as during normal operating conditions. Multiple monitors are considered necessary to cover the ranges of interest.

(1) Noble gas effluent monitors with an upper range capacity of 3.7E+09 M BQ/cc (Xe133) are considered to be practical and should be installed in all operating plants.

(2) Noble gas effluent monitoring shall be provided for the total range of concentration extending from normal condition (as low as reasonably achievable (ALARA)) concentrations to a maximum of 3.7E+09 M Bq/cc (Xe-


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133). Multiple monitors are considered to be necessary to cover the ranges of interest. The range capacity of individual monitors should overlap by a factor of ten.

Because iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radioiodines for the accident condition shall be provided with sampling conducted by absorption on charcoal or other media, followed by onsite laboratory analysis.

In-containment radiation-level monitors with a maximum range of 1E+06 Gy/h shall be installed. A minimum of two such monitors that are physically separated shall be provided. Monitors shall be developed and qualified to function in an accident environment.

A continuous indication of containment pressure shall be provided in the control room of each operating reactor. Measurement and indication capability shall include three times the design pressure of the containment for concrete, four times the design pressure for steel, and –34.32 kPa G for all containments.

A continuous indication of containment water level shall be provided in the control room for all plants. A narrow range instrument shall be provided for BWRs and cover the range from the bottom to the top of the containment sump. A wide range instrument shall also be provided for BWRs and shall cover the range from the bottom of the containment to the elevation equivalent to a 2.27 x 106 liter capacity. For BWRs, a wide range instrument shall be provided and cover the range from the bottom to 1.52 meters above the normal water level of the suppression pool.

A continuous indication of hydrogen concentration in the containment atmosphere shall be provided in the control room. Measurement capability shall be provided over the range of 0 to 10% hydrogen concentration under both positive and negative ambient pressure.

Response

GE believes the requirements of Regulatory Guide 1.97, incorporate the above requirements. Section 7.5 compares the ABWR design against this Regulatory Guide and Subsection 18.8.13 addresses the operator interface of the instrumentation.

1A.2.16 Identification of and Recovery from Conditions Leading to Inadequate Core Cooling [II.F.2]

NRC Position

Licensees shall provide a description of any additional instrumentation controls (primary or backup) proposed for the plant to supplement existing instrumentation (including primary coolant saturation monitors) in order to provide an unambiguous, easy-to-interpret indication of inadequate core cooling (ICC). A description of the


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functional design requirements for the system shall also be included. A description of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided.

Response

The direct water level instrumentation provided in the ABWR design is capable of detecting conditions indicative of inadequate core cooling.

The ABWR has two sets of four wide range reactor water level sensing units (eight total) which are used in two separate two-out-of-four logics which initiate ECCS and other safety functions. Each set of four sensors are used in two separate two-out-of-four logics which initiate ECCS operation. Four separate sets of sensing lines, one from each quadrant of the reactor pressure vessel, supply the pressure to the eight sensors for reliability. The ABWR arrangement of reactor water level sensing complies with the NRC Generic Letter 84-23. The vertical drop inside the drywell of the reactor pressure vessel reference leg water level instrument lines is limited to 0.9 meters. Analog level transmitters are employed to monitor the reactor vessel water level. For the safety related functions initiated automatically upon receipt of a reactor pressure vessel water level trip signal, two-out-of-four trip initiation logic is employed, utilizing a signal from a level transmitter in each of the four instrument divisions. This provides high reliability for initiation upon demand, and high tolerance against inadvertent initiation.

To address the US NRC staff’s concerns about the potential for reactor pressure vessel water level measurement errors resulting from dissolved non-condensable gasses in the water column in the reactor pressure vessel reference leg water level instrument lines (NRC Generic Letter 92-04 and NRC Information Notice 93-27), the ABWR has implemented continuous purging of the reactor pressure vessel reference leg water level instrument lines. Water is continuously injected into the reactor water level reference leg water level instrument lines by means of the Control Rod Drive (CRD) System. This is shown in Figure 5.1-3 and discussed in Subsection 7.7.1.1.

Based on the above information, the existing highly redundant direct water level instrumentation already provides an unambiguous, easy to interpret indication of inadequate core cooling, and there are no plans to include additional instrumentation in the ABWR design. Subsection 18.8.14 addresses the COL license information requirements for these instruments and controls.

1A.2.17 Instruments for Monitoring Accident Conditions [II.F-3]

NRC Position

Provide instrumentation adequate for monitoring plant conditions following an accident that includes core damage.


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Response

The ABWR Standard Plant is designed in accordance with Regulatory Guide 1.97 . A detailed assessment of the Regulatory Guide, including the list of instruments, is found in Section 7.5.

1A.2.18 Safety-Related Valve Position Indication [II.K.1(5)]

NRC Position

(1) Review all valve positions and positioning requirements and positive controls and all related test and maintenance procedures to assure proper ESF functioning, if required.

(2) Verify that AFW valves are in the open position.

Response

(1) The ABWR Standard Plant is equipped with status monitoring that satisfies the requirements of Regulatory Guide 1.47. See Subsection 7.1.2 for detailed information on the status monitoring equipment and capability provided in the ABWR Standard Plant design.

In addition to the status monitoring, the COL applicant plant-specific procedures (Subsection 1A.3.2) will assure that independent verification of system lineups is applied to valve and electrical lineups for all safety-related equipment, to surveillance procedures, to restoration following maintenance and to comply with IE Bulletin 79-08. Through these procedures, approval will be required for the performance of surveillance tests and maintenance, including equipment removal from service and return to service.

(2) This requirement is not applicable to the ABWR. It applies only to Babcock & Wilcox designed reactors.

1A.2.19 Review and Modify Procedures for Removing Safety-Related Systems from Service [II.K.1(10)]

NRC Position

Review and modify (as required) procedures for removing safety-related systems from service (and restoring to service) to assure operability status is known. The COL applicant must verify the operability of safety-related systems after performing maintenance or tests as part of the test to restore a system to service.

Response

See Subsection 1A.3.2 for COL license information requirements.


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1A.2.20 Describe Automatic and Manual Actions for Proper Functioning of Auxiliary Heat Removal Systems When Feedwater System Not Operable [II.K.1(22)]

NRC Position

For boiling water reactors, describe the automatic and manual actions necessary for proper functioning of the auxiliary heat removal systems that are used when the main feedwater system is not operable (Bulletin 79-08, Item 3).

Response

If the main feedwater system is not operable, a reactor scram will be automatically initiated when reactor water level falls to Level 3. The operator can then manually initiate the RCIC System from the main control room, or the system will be automatically initiated as hereinafter described. Reactor water level will continue to decrease due to boiloff until the low-low level setpoint (Level 2) is reached. At this point, the RCIC System will be automatically initiated to supply makeup water to the RPV. This system will continue automatic injection until the reactor water level reaches Level 8, at which time the RCIC steam supply valve is closed.

In the nonaccident case, the RCIC System is normally the only makeup system utilized to furnish subsequent makeup water to the RPV. When the water level reaches Level 2 again due to loss of inventory through the main steam relief valves or to the main condenser, the RCIC System automatically restarts as described in Subsection 1A.2.22. This system then maintains the coolant makeup supply. RPV pressure is regulated by the automatic or manual operation of the main turbine bypass valves which discharge to the condenser.

To remove decay heat during a planned isolation event, assuming that the main condenser is not available, the SRVs are utilized to dump the residual steam to the suppression pool. Suppression pool temperature is monitored by the Suppression Pool Temperature Monitoring (SPTM) System. When the pool temperature increases to a selected setpoint, the SPTM System signals the RHR System to cause automatic initiation of the suppression pool cooling (SPC) mode of RHR. The SPC mode cools the suppression pool by routing pool water through the RHR heat exchanger to cool it, and returning it to the suppression pool. SPC may also be affected by manual alignment of the RHR System. Makeup water to the RPV is still supplied by the RCIC System.

For the accident case with the RPV at high pressure, the HPCF Systems can also be utilized to automatically provide the required makeup flow when the water level drops below Level 1.5 setpoint. No manual operations are required. If the HPCF Systems are postulated to fail at these same conditions and the RCIC capacity is insufficient, the Automatic Depressurization System (ADS) will automatically initiate depressurization of the RPV to permit the low pressure ECCS/LPFL mode of the RHR System to provide makeup coolant.


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Therefore, it can be seen that, although manual actions can be taken to mitigate the consequences of a loss of feedwater, there are no short-term manual actions which must be taken. Sufficient systems exist to automatically mitigate these consequences.

1A.2.21 Describe Uses and Types of RV Level Indication for Automatic and Manual Initiation of Safety Systems [II.K.1(23)]

NRC Position

For boiling water reactors, describe all uses and types of reactor vessel level indication for both automatic and manual initiation of safety systems. Describe other instrumentation that might give the operator the same information on plant status (Bulletin 79-08, Item 4).

Response

The water-level measurement for BWRs, in general, is fully described in NEDO-24708A, “Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors”. An outline of this description as applicable to the ABWR Standard Plant is provided in the following paragraphs.

Figure 7.7-1 illustrates the reactor vessel elevations covered by each water-level range. Additional details may be found in Figure 5.1-3 (Nuclear Boiler System P&ID). The instruments that sense the water level are differential pressure devices calibrated to be accurate at a specific vessel pressure and liquid temperature condition. The following is a description of each water-level range.

(1) Shutdown Water-Level Range—This range is used to monitor the reactor water level during the shutdown condition when the reactor system is flooded for maintenance and head removal. The water-level measurement design is the condensate reference chamber leg type that is not compensated for changes in density. The vessel temperature and pressure conditions that are used for the calibration are 48.9

°C and 0 kPaG water in the vessel. The two vessel instrument penetration elevations used for this water-level measurement are located at the top of the RPV head and the instrument tap just below the bottom of the dryer skirt.

(2) Narrow Water-Level Range—This range uses for its RPV taps the elevation above the main steam outlet nozzle and the tap at an elevation near the bottom of the dryer skirt. The instruments are calibrated to be accurate at the normal operating points. The water-level measurement design is the condensate reference chamber type, is not density compensated, and uses differential pressure devices as its primary elements. The Feedwater Control System uses this range for its water-level control and indication inputs.


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(3) Wide Water-Level Range—This range uses for its RPV taps the elevation above the main steam outlet nozzle and the taps at an elevation near the top of the active fuel. The instruments are calibrated to be accurate at the normal power operating point. The water-level measurement design is the condensate reference type, is not density compensated, and uses differential pressure devices as its primary elements. These instruments provide inputs to various safety systems and engineered safeguards systems.

(4) Fuel-Zone, Water-Level Range—This range used for its RPV taps the elevation above the main steam outlet nozzle and the taps just above the reactor internal pump (RIP) deck. The zero of the instrument is the bottom of the active fuel and the instruments are calibrated to be accurate at 0 MPaG and saturated condition. The water- level measurement design is the condensate reference type, is not density compensated, and uses differential pressure devices as its primary elements. These instruments provide input to water-level indication only.

There are common condensate reference chambers for the narrow-range, wide-range and fuel-zone range water-level transmitters.

The elevation drop from RPV penetration to the drywell penetration is uniform for the narrow-range and wide-range water-level instrument lines in order to minimize the change in water-level with changes in drywell temperature.

Reactor water-level instrumentation that initiates safety systems and engineered safeguards is shown in Figure 5.1-3.

1A.2.21.1 Failure of PORV or Safety to Close [II.K.3.(3)]

NRC Position

Assure that any failure of a PORV or safety valve to close will be reported to the NRC promptly. All challenges to the PORVs safety valves should be documented in the annual report. This requirement is to be met before fuel load.

Response


1A.2.22 Separation of HPCI and RCIC System Initiation Levels [II.K.3(13)]

NRC Position

Currently, the Reactor Core Isolation Cooling (RCIC) System and the High-Pressure Coolant Injection (HPCI) System both initiate on the same low-water-level signal and both isolate on the same high-water-level signal. The HPCI System will restart on low water level but the RCIC System will not. The RCIC System is a low-flow system when compared to the HPCI System. The initiation levels of the HPCI and RCIC Systems


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should be separated so that the RCIC System initiates at a higher water level than the HPCI System. Further, the initiation logic of the RCIC System should be modified so that the RCIC System will restart on low-water-level. These changes have the potential to reduce the number of challenges to the HPCI System and could result in less stress on the vessel from cold water injection. Analyses should be performed to evaluate these changes. The analyses should be submitted to the NRC staff and changes should be implemented if justified by the analyses.

Response

The ABWR Standard Plant design is consistent with this position. The High-Pressure Core Flooder (HPCF) System is initiated at Level one and one half , and the RCIC System is initiated at Level 2. At Level 8, the injection valves for the HPCF and the RCIC steam supply and injection valves will automatically close in order to prevent water from entering the main steamlines.

In the unlikely event that a subsequent low level recurs, the RCIC steam supply and injection valves will automatically reopen to allow continued flooding of the vessel. The HPCF injection valves will also automatically reopen unless the operator previously closed them manually. Refer to Subsections 7.3.1.1.1.1 (HPCF) and 7.3.1.1.1.3 (RCIC) for additional details regarding system initiation and isolation logic.

1A.2.23 Modify Break-Detection Logic to Prevent Spurious Isolation of HPCI and RCIC Systems [II.K.3(15)]

NRC Position

The High-Pressure Coolant Injection (HPCI)and Reactor Core Isolation Cooling (RCIC) Systems use differential pressure sensors on elbow taps in the steamlines to their turbine drives to detect and isolate pipe breaks in the systems. The pipe-break-detection circuitry has resulted in spurious isolation of the HPCI and RCIC Systems due to the pressure spike which accompanies startup of the systems. The pipe-break-detection circuitry should be modified to that pressure spikes resulting from HPCI and RCIC System initiation will not cause inadvertent system isolation.

Response

The ABWR design utilizes the motor-driven HPCF System rather than the turbine-driven HPCI System for high pressure inventory maintenance. Therefore, this position is only applicable to the turbine-driven RCIC System.

In the ABWR Standard Plant design, the high differential pressure signals which isolate the RCIC turbine are processed through the Leak Detection and Isolation System (LDS). Spurious trips are avoided because the RCIC has a bypass start system controlled by valves F037 and F045 (Figure 5.4-8, RCIC P&ID). On receipt of RCIC start signals, bypass valve F045 opens to pressurize the line downstream and accelerate the turbine. The bypass line via F045 is small (25A) and naturally limits the initial flow surge such


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that a differential pressure spike in the upstream pipe will not occur. After a predetermined delay (approximately 5–10 seconds), steam supply valve F037 opens to admit full steam flow to the turbine. At this stage, the line downstream is already pressurized. Thus, it is highly unlikely that a differential pressure spike could occur during any phase of the normal startup process. See Subsection 1A.3.8 for COL license information requirements.

1A.2.24 Reduction of Challenges and Failures of Relief Valves—Feasibility Study and System Modification [II.K.3(16)]

NRC Position

The record of relief-valve failures to close for all boiling water reactors (BWRs) in the past 3 years of plant operation is appproximately 30 in 73 reactor-years (0.41 failures per reactor-year). This has demonstrated that the failure of a relief valve to close would be the most likely cause of a small-break loss-of-coolant accident (LOCA). The high failure rate is the result of a high relief-valve challenge rate and a relatively high failure rate per challenge (0.16 failures per challenge). Typically, five valves are challenged in each event. This results in an equivalent failure rate per challenge of 0.03. The challenge and failure rates can be reduced in the following ways:

(1) Additional anticipatory scram on loss of feedwater

(2) Revised relief-valve actuation setpoints

(3) Increased emergency core cooling (ECC) flow

(4) Lower operating pressures

(5) Earlier initiation of ECC systems

(6) Heat removal through emergency condensers

(7) Offset valve setpoints to open fewer valves per challenge

(8) Installation of additional relief valves with a block or isolation-valve feature to eliminate opening of the safety/relief valves (SRVs), consistent with the ASME Code

(9) Increasing the high steamline flow setpoint for main steamline isolation valve (MSIV) closure

(10) Lowering the pressure setpoint for MSIV Closure

(11) Reducing the testing frequency of the MSIVs

(12) More stringent valve leakage criteria


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(13) Early removal of leaking valves

An investigation of the feasibility and constraints of reducing challenges to the relief valves by use of the aforementioned methods should be conducted. Other methods should also be included in the feasibility study. Those changes which are shown to reduce relief-valve challenges without compromising the performance of the relief valves or other systems should be implemented. Challenges to the relief valves should be reduced substantially (by an order of magnitude).

Response

GE and the BWR Owners’ Group responded to this requirement in Reference 1A-6. This response, which was based on a review of existing operating information on the challenge rate of relief valves, concluded that the BWR/6 product line had already achieved the “order of magnitude” level of reduction in SRV challenge rate. The ABWR relief valve system also has similar design features which also reduce the SRV challenge rate. With regard to inadvertently opened relief valves (IORV), the BWR/6 plant design evaluated for the Owners’ Group report reflected a reduced level if IORV compared with the previous design because of elimination of the pilot-operated relief valve type of design. The ABWR design has also eliminated the pilot-operated relief valve type of design.

For the ABWR, which has a solid-state logic design with redundancy, the likelihood of an IORV is the same or less than the BWR/6 design evaluated in connection with the Owners’ Group report. The redundant-solid state design has been selected in order that the frequency of IORV with solid state-logic becomes low enough so as to achieve the order of magnitude reduction in total SRV challenge rate required by NUREG-0737.

The redundant solid-state design for SRV operation in the pressure relief mode consists of two duplicated microprocessor channels. Each microprocessor channel activates a separate load driver and both load drivers must be activated to cause operation of the SRVs in the relief mode. Operation of the SRVs in the ADS mode also requires activation of two microprocessor channels with separate load drivers to prevent unwanted SRV operation; however, two separate dual channel systems are used to assure reliable operation in the ADS mode. Reliable operation in the pressure relief mode is assured by direct opening of the SRV against spring force.

1A.2.25 Report on Outages of Emergency Core Cooling Systems Licensee Report and Proposed Technical Specification Changes [II.K.3(17)]

NRC Position

Several components of the Emergency Core Cooling (ECC) Systems are permitted by technical specifications to have substantial outage times (e.g., 72 hours for one diesel-generator; 14 days for the HPCI System). In addition, there are no cumulative outage time limitations for ECC Systems. Licensees should submit a report detailing outage


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dates and lengths of outages for all ECC Systems for the last 5 years of operation. The report should also include the causes of the outages (i.e., controller failure, spurious isolation).

Clarification

The present technical specifications contain limits on allowable outage times for ECC Systems and components. However, there are no cumulative outage time limitations on these same systems. It is possible that ECC equipment could meet present technical specification requirements but have a high unavailability because of frequent outages within the allowable technical specifications.

The licensees should submit a report detailing outage dates and length of outages for all ECC Systems for the last 5 years of operation, including causes of the outages. This report will provide the staff with a quantification of historical unreliability due to test and maintenance outages, which will be requirements in the technical specifications.

Based on the above guidance and clarification, a detailed report should be submitted. The report should contain:

(1) Outage dates and duration of outages

(2) Causes of the outage

(3) ECC Systems or components involved in the outage

(4) Corrective action taken

Tests and maintenance outages should be included in the above listings which are to cover the last 5 years of operation. The licensee should propose changes to improve the availability of ECC equipment, if needed.

Applicants for an operating license shall establish a plan to meet these requirements.

Response


1A.2.26 Modification of Automatic Depressurization System Logic—Feasibility for Increased Diversity for Some Event Sequences [II.K.3(18)]

NRC Position

The Automatic Depressurization System (ADS) actuation logic should be modified to eliminate the need for manual actuation to assure adequate core cooling. A feasibility and risk assessment study is required to determine the optimum approach. One possible scheme that should be considered is ADS actuation on low reactor-vessel water level, provided no HPCI or HPCS flow exists and a low-pressure ECC System is running. This logic would complement, not replace, the existing ADS actuation logic.


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Response

An 8 minute high drywell pressure bypass timer has been added to the ADS initiation logic to address TMI action Item II.K.3.18. This timer will initiate on a Low Water Level 1 signal. When it times out, it bypasses the need for a high drywell signal to initiate the standard ADS initiation logic.

For all LOCAs inside the containment, a high drywell signal will be present and ADS will actuate 29 seconds after a Low Water Level 1 signal is reached. All LOCAs outside the containment become rapidly isolated and any one of the three high pressure ECCS can control the water level. The high drywell pressure bypass timer in the ADS initiation logic will only affect the LOCA response if all high pressure ECCS fail following a break outside the containment. For this case, the ADS will automatically initiate within 509 seconds (8 minute timer plus 29 second standard ADS logic delay) following a Low Water Level 1 signal.

1A.2.27 Restart of Core Spray and LPCI Systems on Low Level Design and Modification [II.K.3(21)]

NRC Position

The Core Spray and Low Pressure Coolant Injection (LPCI) Systems flow may be stopped by the operator. These systems will not restart automatically on loss of water level if an initiation signal is still present. The Core Spray and LPCI system logic should be modified so that these systems will restart, if required, to assure adequate core cooling. Because this design modification affects several core-cooling modes under accident conditions, a preliminary design should be submitted for staff review and approval prior to making the actual modification.

Response

The ABWR Standard Plant Emergency Core Cooling System (ECCS) is made up of the High Pressure Core Flooder (HPCF) System, the Reactor Core Isolation Cooling (RCIC) System, the Automatic Depressurization System (ADS) and the low pressure flooder (LPFL) mode of the Residual Heat Removal (RHR) System.

A high water level (Level 8) signal will automatically close the HPCF injection valves and the RCIC steam supply and injection valves. These systems may also be shut down manually. Manual action is required to shut down the RHR once it is initiated.

In the unlikely event that a subsequent low level reoccurs, the RCIC steam supply and injection valves will automatically reopen to allow continued flooding of the vessel. The HPCF injection valves will also automatically reopen unless the operator previously closed them manually.

The NRC has suggested certain modifications to the LPCI (LPFL for the ABWR) and core flooder systems provided as part of the ECCS network. These NRC suggestions


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center on control system logic modifications that would provide automatic restart capability following manual termination of system operation.

General Electric and the BWR Owners’ Group reviewed this issue on a generic basis in 1980, and concluded that the NRC suggestions were not required for plant safety considerations. Justification is provided in the December 29, 1980, BWR Owners’ Group submittal to the NRC (Reference 1A-8). Plant variations between the BWR and the ABWR designs are not significant to the overall technical conclusions of the study.

This conclusion is based on the adequacy of the current ECCS logic design coupled with the potentially negative impact on overall safety of the proposed changes. For the low pressure ECCS, these negative impacts include a significant escalation of control system complexity and restricted operator flexibility when dealing with anticipated events.

A full understanding of the significance of these logic changes must be based on a recognition that these systems must consider the possible interactive effects among the other systems making up the overall ECCS network. This must also include the potential impact on supporting systems such as the standby power supplies and the shutdown service water systems. Furthermore, the LPFL is one mode of the RHR System which has other safety-related functions such as suppression pool cooling and wetwell/drywell spray cooling. Clearly, these other safety functions must not be compromised by any changes in the LPFL mode of operation.

The referenced systems analysis took into consideration these potential interactive effects, impacts on supporting systems, plant instrumentation and emergency procedure guidelines. The study concluded that auto-restart of these systems is not necessary or appropriate. Therefore, GE concludes that no modifications should be made with respect to automatic restart of the low pressure ECCS.

1A.2.28 Automatic Switchover of Reactor Core Isolation Cooling System Suction—Verify Procedures and Modify Design [II.K.3(22)]

NRC Position

The Reactor Core Isolation Cooling (RCIC) System takes suction from the condensate storage tank (CST) with manual switchover to the suppression pool when the CST level is low. This switchover should be made automatically. Until the automatic switchover is implemented, licensees should verify that clear and cogent procedures exist for the manual switchover of the RCIC System suction from the condensate storage tank to the suppression pool.

Response

The RCIC System provided in the ABWR Standard Plant includes an automatic switchover feature which will change the pump suction source from the condensate storage pool to the suppression pool. The safety-grade switchover will automatically


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occur upon receipt of a low-level signal from the condensate storage tank or a high-level signal from the suppression pool.

See Subsection 7.3.1.1.1.3 for additional information.

1A.2.29 Confirm Adequacy of Space Cooling for High Pressure Coolant Injection and Reactor Core Isolation Cooling Systems [II.K.3(24)]

NRC Position

Long-term operation of the Reactor Core Isolation Cooling (RCIC) and High-Pressure Coolant Injection (HPCI) Systems may require space cooling to maintain the pump-room temperatures within allowable limits. Licensees should verify the acceptability of the consequences of a complete loss of alternating-current power. The RCIC and HPCI Systems should be designed to withstand a complete loss of offsite alternating-current power to their support systems, including coolers, for at least 2 hours.

Response

The ABWR HPCF and the RCIC systems are provided space cooling via individual room safety grade air-conditioning systems (Subsection 9.4.5). If all offsite power is lost, space cooling for the HPCF and RCIC System equipment would not be lost because the motor power supply for each system is from separate essential power supplies.

1A.2.30 Effect of Loss of Alternating Current Power on Pump Seals [II.K.3(25)]

NRC Position

The licensees should determine, on a plant-specific basis, by analysis or experiment, the consequences of a loss of cooling water to the reactor recirculation pump seal coolers. The pump seals should be designed to withstand a complete loss of alternating-current (AC) power for at least 2 hours. Adequacy of the seal design should be demonstrated.

Response

The ABWR design features internal recirculation pumps called Reactor Internal Pumps (RIP) which do not require shaft seals. During a Loss of Preferred Power (LOPP), the RIPs shutdown automatically but there are no shaft seals which require cooling water restoration.

A plant AC power failure would temporarily disrupt the operation of the reactor recirculation subsystems but their failure would not generate a LOCA, as the following describes.

(1) RMC Failure—Subsection 5.4.1.3.1 describes the RMC operation during normal running or stopped condition. This operation assumes that the Reactor Building Cooling Water (RCW) System is in operation continuously during these operating or stopped conditions. Normal LOCA and LOPP operation of the RCW are described in Subsection 9.2.11.2.


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A loss of AC power or Loss of preferred power (LOPP) will stop all RIPs. The LOPP will also temporarily stop the RCW and RSW pumps. The onsite emergency power sources will automatically restart the RCW/RSW pumps, which will restore cooling for the stopped RIP motors. The RCW primary containment isolation valves will not close on LOPP (only on LOCA).

The RMC Subsystem for each RIP includes a motor cooling inlet and outlet temperature detectors TE 301 and 302, which will automatically run back individual RIPs to minimum speed and subsequently trip on high coolant temperature and prevent motor damage.

The RIP motors are designed and will be plant tested to not be damaged in the stopped hot standby condition indefinitely with RCW cooling available.

(2) RMP Failure—Subsection 5.4.1.3.2 describes the RMP operation. Since the RIP and motor have no seals, the water in the RIP motor communicates directly with the reactor water at the same pressure but at much lower temperature. There is no possibility of this water escaping from the coolant pressure boundary such as in conventional pumps, which include seals.

The RMP water is supplied from the CRD System. The CRD pumps will stop temporarily during a LOPP, which will cause the normal RMP flow to each RIP to temporarily stop. The CRD pumps are subsequently restarted automatically, after several minutes time delay, powered by onsite power sources and the RMP water will restart.

This temporary interruption of RMP flow will not initiate a LOCA. The only effect of loosing the RMP flow temporarily to the RIPs, from a LOPP, is that it will allow reactor contamination, by diffusion, to enter the RIP motor during the RMP flow interruption.

(3) RMISS Failure—Subsection 5.4.1.3.3 describes the operation of the RMISS, which is used only during plant shutdown and RIP maintenance. The power source for the inflatable seal is a portable air-operated water pump which is moved from RIP to RIP. A LOPP would therefore not cause a direct loss of RMISS pressure, since (1) the plant air system has a finite passive storage capacity in the air receivers, and (2) the RMISS air-operated pump only operates when the RMISS pressure drops below a preset value.

The RMISS is a secondary seal. Even with a long time RMISS failure RIP maintenance, the passive backseat seal of the RIP shaft on the stretch tube will preclude draining the reactor.


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1A.2.31 Study and Verify Qualification of Accumulators on Automatic Depressurization System Valves [II.K.3(28)]

NRC Position

Safety analysis reports claim that air or nitrogen accumulators for the Automatic Depressurization System (ADS) valves are provided with sufficient capacity to cycle the valves open five times at normal drywell pressures. GE has also stated that the Emergency Core Cooling (ECC) Systems are designed to withstand a hostile environment and still perform their function for 100 days following an accident. The licensee should verify that the accumulators on the ADS valves meet these requirements, even considering normal leakage. If this cannot be demonstrated, the licensee must show that the accumulator design is still acceptable.

Response

The accumulators for the ADS valves are sized to provide one actuation at drywell design pressure and five actuations at normal drywell pressure. This cyclic capability is validated during preoperational testing at the station. The accumulators are safety-grade components.

The 100-day, post-accident functional operability requirement is met through conservative design and redundancy; eight ADS valves are provided with code-qualified accumulators and Seismic Category I piping within primary containment. Two redundant seven-day supplies of bottled air are available to compensate for leakage during long-term usage, with replacement capability being provided for the remainder of the postulated accident to assure system functional operability. A maximum of three of eight ADS valves need to function to meet short-term demands (Subsection 19.3.1.3.1) and the functional operability of only one ADS valve will fulfill longer term needs. Loss of pneumatic supply pressure to the ADS SRV accumulator is alarmed to provide the reactor operator with indication of the failure of any of the redundant systems under hostle environmental conditions.

The BWR Owners’ Group sponsored an evaluation of the adequacy of the ADS configurations. Evaluation results are summarized in the following paragraph.

The accumulators are designed to provide one actuation at drywell design pressure and five actuations at normal drywell pressure. See Subsection 6.7.1 for a description of the ADS N2 pneumatic supply.

1A.2.32 Revised Small-Break Loss-of-Coolant-Accident Methods to Show Compliance with 10CFR50, Appendix K [II.K.3(30)]

NRC Position

The analysis methods used by Nuclear Steam Supply System (NSSS) vendors and/or fuel suppliers for small break loss-of-coolant accident (LOCA) analysis for compliance with Appendix K to 10CFR50 should be revised, documented, and submitted for NRC


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approval. The revisions should account for comparisons with experimental data, including data from the LOFT Test and Semiscale Test facilities.

Response

GE has evaluated the NRC request requiring that the BWR small-break LOCA analysis methods are to be demonstrated to be in compliance with Appendix K to 10CFR50 or that they be brought into compliance by analysis methods changes. The specific NRC concerns are contained in NUREG-0626, Appendix F. The specific NRC concerns identified in Subsection 4.2.10 of NUREG-0626 (Appendix F) relate to the following: counter current flow limiting (CCFL) effects, core bypass modeling, pressure variation in the reactor pressure vessel, integral experimental verification, quantification of uncertainties in predictions, the recirculation line inventory modeling, and the homogeneous/equilibrium model.

The response to the NRC small-break model concerns was provided at a meeting between the NRC and GE on June 18, 1981. Information provided at this meeting showed that, based on the TLTA small-break test results and sensitivity studies, the existing GE small-break LOCA model already satisfies the concerns of NUREG-0626 and is in compliance with 10CFR50, Appendix K. Therefore, the GE model is acceptable relative to the concerns of Item II.K.3(30), and no model changes need be made to satisfy this item.

Documentation of the information provided at the June 18, 1981 meeting was provided via the letter from R. H. Buchholz (GE) to D. G. Eisenhut (NRC), dated June 26, 1981.

1A.2.33 Plant-Specific Calculations to Show Compliance with 10CFR50.46 [II.K.3(31)]

NRC Position

Plant-specific calculations using NRC-approved models for small break loss-of-coolant accidents (LOCAs) as described in Item II.K.3.30 to show compliance with 10CFR50.46 should be submitted for NRC approval by all licensees.

Response

The ABWR standard safety small-break LOCA calculations are discussed in Subsection 6.3.3.7.

The references listed in Subsection 6.3.6 describe the currently approved Appendix K methodology used to perform these calculations. Compliance with 10CFR50.46 has previously been established for that methodology.

Since, as noted in the previous Item (1A.2.32), no model changes are needed to satisfy NUREG-0737, Item II.K.3(30), there is no need to revise the calculations presented in Subsection 6.3.3.7.


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1A.2.33.1 Evaluation of Anticipated Transients with Single Failure to Verify No Fuel Failure [II.K.3 (44)]

NRC Position

For anticipated transients combined with the worst single failure and assuming proper operator actions, licensees should demonstrate that the core remains covered or provide analysis to show that no significant fuel damage results from core uncovery. Transients which occur in a stuck-open relief valve should be included in this category. The results of the evaluation are due January 1, 1981.

Response

GE and the BWROG have concluded, based on a representative BWR/6 plant study, that all anticipated transients in Regulatory Guide 1.70, Revision 3, combined with the worst single failure, the reactor core remains covered with water until stable conditions are achieved. Furthermore, even with more degraded conditions involving a stuck-open relief valve in addition to the worst transient (loss of feedwater) and worst single failure (of high pressure core spray), studies show that the core remains covered and adequate core cooling is available during the whole course of the transient (NEDO-24708, March 31, 1980). The conclusion is applicable to the ABWR. Since the ABWR has more high pressure makeup systems (2HPCFs and 1 RCIC), the core covering is further assured.

Other discussions of transients with single failure is presented in the response to NRC Question 440.111.

1A.2.33.2 Evaluate Depressurization Other Than Full ADS [II.K.3 (45)]

NRC Position

Provide an evaluation of depressurization methods, other than by full actuation of the Automatic Depressurization System, that would reduce the possibility of exceeding vessel integrity limits during rapid cooldown (Applicable to BWRs only).

Response

This response is provided in Subsection 19A.2.11.

1A.2.33.3 Responding to Michelson Concerns [II.K.3 (46)]

NRC Position

General Electric should provide a response to the Michelson concerns as they relate to boiling water reactors.

Clarification: General Electric provided a response to the Michelson concerns as they relate to boiling water reactors by letter dated February 21, 1980. Licensees and applicants should assess applicability and adequacy of this response to their plants.


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Response

All of the generic February 21, 1980 GE responses were reviewed and updated for the ABWR Standard Plant. The specific responses are provided in Table 1A-1.

1A.2.34 Primary Coolant Sources Outside Containment Structure [III.D.1.1(1)]

NRC Position

Applicants shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels. This program shall include the following:

(1) Immediate Leak Reduction

(a) Implement all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment

(b) Measure actual leakage rates with systems in operation and report them to the NRC

(2) Continuing Leak Reduction—establish and implement a program of preventive maintenance to reduce leakage to as-low-as-practical levels. This program shall include periodic integrated leak tests at intervals not to exceed each refueling cycle.

Response

Leak reduction measures of the ABWR Standard Plant include a number of barriers to containment leakage in the closed systems outside the containment. These closed systems include:

(1) Residual Heat Removal

(2) High Pressure Core Flooder

(3) Low Pressure Core Flooder

(4) Reactor Core Isolation Cooling

(5) Suppression Pool Cleanup

(6) Reactor Water Cleanup

(7) Fuel Pool Cooling and Cleanup

(8) Post-Accident Sampling

(9) Process Sampling

(10) Containment Atmospheric Monitoring

(11) Fission Product Monitor (Part of LDS)

(12) Hydrogen Recombiner

(13) Standby Gas Treatment


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Plant procedures will prescribe the method of leak testing these systems. The testing will be performed on a schedule appropriate to 10CFR50 Appendix J type B and C penetrations (i.e., at each refueling outage). When leakage paths are discovered, including during these tests, they will be investigated and necessary maintenance will be performed to reduce leakage to its lowest practical level.

In addition, lines which penetrate the primary containment are equipped with inboard and outboard isolation valves that are designed in accordance with GDC 55, 56 and 57 to provide reliable isolation in the event of a line break or leakage. The containment isolation provisions are discussed in detail in Subsection 6.2.4, which also identifies all the system lines that penetrate the containment, together with their respective isolation valves.

Leakage within and outside the primary containment are continuously monitored by the Leak Detection and Isolation System (LDS) for breach in the integrity of the containment. Upon detection of a leakage parameter, the LDS will automatically initiate the necessary control functions to isolate the source of the break and alerts the operator for corrective action. The MSL tunnel area is monitored for high radiation levels and for high ambient temperatures that are indicative of steam leakage. The Turbine Building is also monitored for high area ambient temperatures for MSL leakage. The resulting action causes isolation of the MSIVs and subsequent shutdown of the reactor.

The radiation levels in the HVAC air exhaust ducts of the Reactor Building and the fuel handling area are monitored for use in isolating the secondary containment. The results in closure of the HVAC air ducts in the Reactor Building, closure of the containment purge and vent lines, and activation of the Standby Gas Treatment System (SGTS).

The leak detection methods and associated instrumentation that monitor leakage from the reactor coolant pressure boundary are described in Subsection 5.2.5.

For small line breaks in the secondary containment that could cause significant release of radioactive material will be detected by process radiation monitors in the Reactor Building HVAC air ducts. As indicated above, a high radiation level will activate the SGTS (Subsection 6.5.1) prior to the release of radiation to the environment. Also, any fluid leakage will drain into the sumps in the Reactor Building and will be monitored by sumps instrumentation for level and flow rate. The operator will be alerted to any abnormal condition for action. All lines which pass outside of the secondary containment contain leakage detection systems or loop seals. These systems allow the SGTS to maintain a negative pressure relative to the environment and thus limit the amount of leakage through the secondary containment. These systems are discussed in Subsection 6.2.3. Finally, expected liquid leakoff from equipment outside the containment is directed to equipment drain sumps and processed by the Radwaste


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System. These multiple design features of the ABWR Standard Plant provide substantial capability to limit any potential release to the environment from systems likely to contain radioactive material.

Additionally, pressure boundary components of radioactive waste systems are purchased as augmented Class D systems to assure their capability to provide integrity.

1A.2.35 In-Plant Radiation Monitoring [II.D.3.3(3)]

NRC Position

(1) Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident.

(2) Each applicant for a fuel-loading license to be issued prior to January 1, 1981 shall provide the equipment, training, and procedures necessary to accurately determine the presence of airborne radioiodine in areas within the plant where plant personnel may be present during an accident.

Response

(1) See Subsection 1A.3.3 for COL license information requirements.

(2) Not applicable.

1A.2.36 Control Room Habitability [III.D.3.4(1)]

NRC Position

In accordance with Task Action Plan Item III.D.3.4 and control room habitability, licensees shall assure that control room operators will be adequately protected against the effects of accidental release of toxic and radioactive gases and that the nuclear power plant can be safely operated or shut down under design basis accident conditions (Criterion 19, “Control Room” of Appendix A, General Design Criteria for Nuclear Power Plants, to 10CFR50).

Response

This requirement is satisfied for the ABWR by the provisions of the Control Building Atmospheric Control System(ACS).

Section 7.1 describes the Control Building ACS instrumentation and controls for assuring control room habitability. Subsection 6.4 provides HVAC design details. As demonstrated by the analyses provided in these subsections, these systems satisfy Criterion 19, Appendix A of 10CFR50.


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The ABWR design satisfies this item.

1A.3 COL License Information

1A.3.1 Emergency Procedures and Emergency Procedures Training Program

Emergency procedures, developed from the emergency procedures guidelines, shall be provided and implemented prior to fuel loading (Subsection 1A.2.1).

1A.3.2 Review and Modify Procedures for Removing Safety-Related Systems from Service

Procedures shall be reviewed and modified (as required) for removing safety-related systems from service (and restoring to service) to assure operability status is known (Subsections 1A.2.18 and 1A.2.19).

1A.3.3 In-Plant Radiation Monitoring

Equipment and training procedures shall be provided for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during the accident (Subsection 1A.2.35).

1A.3.4 Reporting Failures of Reactor System Relief Valves

Failures of reactor system relief valves shall be reported in the annual report to the NRC (Subsection 1A.2.21.1).

1A.3.5 Report on ECCS Outages

Starting from the date of commercial operations, an annual report should be submitted which includes instance of ECCS unavailability because of component failure, maintenance outage (both forced or planned), or testing, the following information shall be collected:

(1) Outage date

(2) Duration of outage

(3) Cause of outage

(4) Emergency core cooling system or component involved

(5) Corrective action taken

The above information shall be assembled into a report, which will also include a discussion of any changes, proposed or implemented, deemed appropriate, to improve the availability of the emergency core cooling equipment (Subsection 1A.2.25).


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1A.3.6 Procedure for Reactor Venting

Procedures shall be developed for the operators use of the reactor vents. (See Subsection 1A.2.5)

1A.3.7 Testing of SRV and Discharge Piping

The COL applicant will confirm that any SRVs or discharge piping installed that is not similar to those that have been tested will be tested in accordance with Subsection 1A2.9.

1A.3.8 RCIC Bypass Start System Test

The COL applicant shall perform the RCIC bypass start system test described in Subsection 1A.2.23 during plant startup.

1A.4 References

1A-1 Memo, C. Michelson to D. Okrent, “Possible Incorrect Operator Action Such as Pipe Break Isolation”, June 4,1979.

1A-2 Letter, D. G. Eisenhut (NRC)to R. L. Gridley(GE), “Potential for Break Isolation and Resulting GE-Recommended BWR/3 ECCS Modifications”, June 14, 1978.

1A-3 “Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors”, NEDO-24708, August 1979.

1A-4 U. S. Nuclear Regulatory Commission, “NRC Action Plan Developed as a Result of the TMI-2 Accident”, USNRC report NUREG-0660, Vols. 1 and 2, May 1980.

1A-5 U. S. Nuclear Regulatory Commission, “Clarification of TMI Action Plan Requirements”, USNRC Report NUREG-0737, November 1980.

1A-6 U. S. Nuclear Regulatory Commission, “Licensing Requirements for Pending Applications for Construction Permits and Manufacturing License”, NUREG-0718, Revision 1, June 1981.

1A-7 Letter, R. H. Buchholz (GE) to D. F. Ross (NRC), Subject: Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors, November 30, 1979, MFN-290-79.

1A-8 Letter, D. B. Waters (Chairman, BWR Owners’ Group) to NRC staff, dated December 29, 1980, “BWR Owners’ Group Evaluation of NUREG-0737 Requirements.”


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1A-9 Letter, D. B. Waters (Chairman, BWR Owners’ Group) to D. G. Eisenhut (NRC), dated March 31, 1981, “BWR Owners’ Group Evaluation of NUREG-0737 Requirements II.K.3.16 and II.K.3.18.”

1A-10 Letter, D.B.Waters (Chairman, BWR Owners’ Group) to R.H. Vollmer (NRC), dated September 17, 1980, NUREG-0578 “Requirements 2.12-Performance Testing of BWR and PWR Relief and Safety Valves.”

1A-11 NEDE-24988-P, “Analysis of Generic BWR Safety/Relief Valve Operability Test Results”, Proprietary Document, October 1981.


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Table 1A-1 Responses to Questions Posed by Mr. C. Michelson [II.K.3(46)]

Question 1

Pressurizer level is an incorrect measure of primary coolant inventory.

Response 1

BWRs do not have pressurizers. BWRs measure primary coolant inventory directly using differential pressure sensors attached to the reactor vessel. Thus, this concern does not apply to the ABWR.

Question 2

The isolation of small breaks (e.g., letdown line; PORV) not addressed or analyzed.

Response 2

Automatic isolation only occurs for breaks outside the containment. Such breaks are addressed in Section 3.1.1.1.2 of NEDO-24708. It was shown that if high pressure systems are available, no operator actions are required. If it is assumed that all high pressure systems fail, the operator must manually depressurize to allow the low pressure systems to inject and maintain vessel water level. Analyses in Section 3.5.2.1 of NEDO-24708 show that the operator has sufficient information and time to perform these manual actions. The necessary manual actions have been included in the operator guidelines for small-break accidents.

Question 3

Pressure boundary damage due to loadings from (1) bubble collapse in subcooled liquid and 2) injection of ECC water in steam-filled pipes.

Response

The BWR has no geometry equivalent to that identified in Michelson’s report on B&W reactors relative to bubble collapse (steam bubbling upward through the pressurizer surge line and pressurizer). Thus, the first concern is not applicable to the ABWR.

ECC injection in the ABWR at high pressure is either into the reactor vessel throuh water-filled lines (RHR-B+C;HPCF-B+C) or into the feedwater lines (RHR-A:RCIC). The feedwater lines are normally filled with relatively cold liquid (251.5oC or less). ECCS injection in the ABWR at low pressure is either directly into the reactor vessel or into the feedwater lines. Thus, the second concern is not applicable to the ABWR.

Question 4

In determining need for steam generators to remove decay heat, consider that break flow enthalpy is not core exit enthalpy.

Response 4

BWRs do not use steam generators to remove decay heat, so this concern does not apply to the ABWR.

Question 5

Are sources of auxiliary feedwater adequate in the event of a delay in cooldown subsequent to a small LOCA?


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Response 5

BWRs do not need feedwater to remove heat from the reactor following a LOCA, whether the subsequent cooldown is delayed or not. Therefore, this concern is not applicable to the ABWR. BWRs have a closed cooling system in which vessel water flows out the postulated break to the suppression pool. The suppression pool is cooled and water is pumped back to the vessel with ECCS pumps.

Question 6

Is the recirculation mode of operation of the HPCI pumps at high pressure an established design requirement?

Response 6

The high-pressure injection systems utilized in the ABWR are the Reactor Core Isolation Cooling (RCIC) and High Pressure Core Flooder (HPCF) Systems.

The RCIC and HPCF Systems normally take suction from the condensate storage tank and have an alternate suction source from the suppression pool. A recirculation mode of operation of these systems is established when the system suction is from the suppression pool. Following a LOCA when system suction is from the suppression pool, water injected into the reactor is discharged through the break and flows back to the suppression pool, forming a closed recirculation loop.

Other recirculation modes include test modes (e.g., suction from and discharge to the suppression pool) and system operation on low flow bypass with discharge to the suppression pool.

All of these modes are established design requirements.

Question 7

Are the HPCI pumps and RHR pumps run simultaneously? Do they share common piping?/suction? If so, is the system properly designed to accommodate this mode of operation (i.e., are any NPSH requirements violated, etc...)?

Response 7

For the ABWR, the high-pressure injection systems (RCIC/HPCF) do not share any common suction piping with the low pressure RHR and they can operate simultaneously with this low pressure system.

The RCIC and HPCF Systems share common suction line from the condensate storage tank. The RHR shutdown cooling operating mode does not share any common suction piping with the RCIC or HPCF Systems. It is an established design requirement to size the suction piping, including shared piping, such that adequate NPSH is available to RCIC and HPCF pumps for all operating modes of these systems.

Pre-operational and/or startup tests are conducted that demonstrate that the NPSH requirements are met.

Question 8

Mechanical effects of slug flow on steam generator tubes need to be addressed (transitioning from solid natural circulation to reflux boiling and back to solid natural circulation may cause slug flow in the hot leg pipes).


(Continued)


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Response 8

BWRs do not have steam generators, so this concern does not apply to ABWR.

Question 9

Is there minimum flow protection for the HPCI pumps during the recirculation mode of operation?

Response 9

For the ABWR, the RCIC, RHR, and HPCF pumps all contain valves, piping, and automatic logic that bypasses flow to the suppression pool as required to provide minimum flow protection for all design basis operating modes of the systems.

Question 10

The effect of the accumulators dumping during small-break LOCAs is not taken into account.

Response 10

BWRs do not use accumulators to mitigate LOCAs. Therefore, this concern does not apply to the ABWR.

Question 11

What is the impact of continued running of the RC pumps during a small LOCA?

Response 11

The impact of continued running of the recirculation pumps has been addressed in Subsections 3.3.2.2, 3.3.2.3, and Subsection 3.5.2.1.5.1 of NEDO-24708. The conclusions were that continued running of the recirculation pumps results in little change in the time available for operator actions and does not significantly change the overall system response.

Question 12

During a small break LOCA in which offsite power is lost, the possibility and impact of pump seal damage and leakage has not been evaluated.

Response 12

The RCIC, HPCF, and RHR pumps are provided with mechanical seals. No external cooling from auxiliary support systems, such as site service water or room air coolers, is required for RCIC pump seals. The HPCF and RHR seals are cooled by connections to the three separate divisions of the Reactor Building Cooling Water (RCW) System to protect against single failures. RHR Divisions A, B and C, and HPCF-B and C are connected to their corresponding RCW divisions. If offsite power is lost, onsite diesel generators maintain the RCW three-divisional function. These types of seals have demonstrated (in nuclear and other applications) their capability to operate for an extended period of time at temperatures in excess of those expected following a LOCA.

Should seal failure occur, it can be detected by room sump high level alarms. The RCIC, HPCF, and RHR individual pumps are arranged, and motor-operated valves provided, so that a pump with a failed seal can be shut down and isolated without affecting the proper operation of the other redundant pumps/systems.


(Continued)


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Considering the low probability of seal failure during a LOCA, the fact that a pump with a failed seal can be isolated without affecting other redundant equipment, and the substantial redundancy provided in the BWR emergency cooling systems, pump seal failure is not considered a significant concern.

Question 13

During transitioning from solid natural circulation to reflux boiling and back again, the vessel level will be unknown to the operators, and emergency procedures and operator training may be inadequate. This needs to be addressed and evaluated.

Response 13

There is no similar transition in the BWR case. In addition, the BWR has water level measurement within the vessel and the indication of the water level is incorporated into the operator guidelines. Consequently, this concern does not apply to ABWR.

Question 14

The effect of non-condensable gas accumulation in the steam generators and its possible disruption of decay heat removal by natural circulation needs to be addressed.

Response 14

The effect of non-condensable gas accumulation is addressed in Subsection 3.3.1.8.2 of NEDO-24708. For a BWR, vapor is present in the core during both normal operation and natural circulation conditions. Non-condensables may change the composition of the vapor but would have an insignificant effect on the natural or forced circulation itself, since the non-condensables would rise with the steam to the top of the vessel after leaving the steam separators.

Concern 15

Delayed cooldown following a small-break LOCA could raise the containment pressure and activate the containment spray system.

Response 15

The ABWR containment spray system is manually initiated. All essential equipment inside the containment is required to be qualified for the environmental conditions resulting from the initiation of the containment spray system.

Question 16*

This concern relates to the possibility that an operator may be included and perhaps even trained to isolate, where possible, a pipe break LOCA without realizing that it might be an unsafe action leading to high pressure, and short-term core bakeout. For example, if a BWR should experience a LOCA from a pressure boundary failure somewhere between the pump suction and discharge valve for either reactor recirculation pump, it would be possible for the operator to close these valves following the reactor blowdown to repressurize the reactor coolant system. Before such isolation should be permitted, it is first necessary to show by an appropriate analysis that the high pressure ECCS is adequate to reflood the uncovered core without assistance from the low pressure ECCS, which can no longer deliver flow because of the repressurization. Otherwise, such isolation action should be explicitly forbidden in the emergency operating instructions.


(Continued)


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* Excerpt from Reference 1A-1

Response 16

The ABWR does not have recirculation lines. However, there are other systems where it is possible for the operator to close these valves following the reactor blowdown to low pressure and thereby isolate the break. An example of this would be a break in the reactor water cleanup piping between the shutdown suction line valve and the containment boundary. In Reference 1A-2, the NRC concluded that, based on information provided by GE, break isolation is not a problem.

In order for the reactor vessel to repressurize following isolation of a line break, the isolation would have to occur before initiation of ADS due to a high drywell pressure in concurrence with low water Level 1 condition. Isolation of a line break prior to obtaining a high drywell pressure signal might occur for very small breaks (area << 9.3 cm2), which may require several hundred seconds following the break to reach the high drywell pressure setpoint. In this case it has been shown in Reference 1A-3 that the high pressure systems (RCIC, HPCF and feedwater) are sufficient to maintain the water level above the top of the core.

If isolation of the break were to occur prior to reaching Level 1 but after the high drywell pressure signal, the vessel would pressurize to the SRV setpoint following isolation of the main steamlines and then oscillate as the SRVs cycle opened and closed. If no high pressure systems were available, the loss of mass out the SRVs would cause the level to continue dropping and result in automatic ADS actuation shortly after reaching Level 1. This would depressurize the vessel and allow the low pressure systems to begin injecting. This capability was demonstrated in NEDO-24708, explicitly to provide for manual depressurization in the event of low reactor water level with high pressure systems unable to maintain level for any reason.

In summary, in order to repressurize the vessel following break isolation, the isolation would have to occur prior to ADS blowdown. For these cases, high pressure systems would maintain inventory. If no high pressure system was available, the low pressure systems would control the vessel water level following automatic or manual vessel depressurization.


(Continued)

Plant Shielding to Provide Access to Vital Areas 1AA-1

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1AA Plant Shielding to Provide Access to Vital Areas and Protective Safety Equipment for Post-Accident Operation [II.B.2]

1AA.1 Introduction

General Electric has performed a review of the ABWR Standard Plant post-accident environment in response to NUREG-0737 Item II.B.2. This attachment discusses the results of that review.

1AA.2 Summary of Shielding Design Review

Several alternatives are potentially available to the designer to assure continued equipment availability and performance under post-accident conditions. One is to provide redundant systems and/or components which are qualified to operate in the expected environment. Another is to provide operator access to conduct the operations and to maintain the equipment. This latter alternative would generally be accompanied by appropriate shielding and in many cases would be difficult if not impossible to carry out.

GE has taken the first approach and furthermore has designed the plant so that most responses to transient conditions are automatic, including achieving and maintaining safe-shutdown conditions. The design basis for the ABWR Standard Plant is to require safety-related equipment to be appropriately environmentally qualified and operable from the control room. As a result of this design philosophy and as shown by this review, no changes are necessary to assure that personnel access is adequate or that safety equipment is not degraded because of post-accident operation.

As part of the design of the ABWR Standard Plant, it was necessary to establish the environmental conditions for qualification of safety-related equipment. A result of this design work was an environmental requirement establishing the integrated dose that the equipment must be able to withstand. These values are listed in Appendix 3I.

Another aspect of the review was the manner in which the safety-related equipment is arranged and operated during normal and abnormal operation and postulated accidents. The essence of the ABWR Standard Plant is to achieve and maintain a safe shutdown condition for all postulated accident conditions with all operator actions being conducted from outside the primary and secondary containment zones, principally from the control room.

The purpose of this review is first to verify that, where equipment access is required, it is reasonably accessible outside the primary and secondary containment zones. Secondly, the review should verify that inaccessible equipment is environmentally qualified and is operable from the control room.

1AA-2 Plant Shielding to Provide Access to Vital Areas

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The results of the review are:

(1) The period of interest begins with the plant in a safe shutdown condition. Thus, the various safety-related systems needed to achieve safe shutdown conditions have performed as expected, and only the engineered safety features systems (Chapter 6) and auxiliaries, as described later, are required to maintain this condition.

(2) Based upon the accident source terms of Regulatory Guides 1.3 and 1.7 and Standard Review Plan 15.6.5 including normal operations, the vital equipment exposures will be enveloped based upon the table below:

Each actual area will be environmentally qualified to the area specific envelope as defined in Tables 3I.3-9 through 3I.3-13 and 3I.3-19 through 3I.3-20.

(3) It is not necessary for operating personnel to have access to any place other than the control room, technical support center, post-accident sampling station, sample analysis area, and safety-related nitrogen supply bottles to operate the equipment of interest during the 100-day period. The control room, technical center and sample analysis area are designed to be accessible post-accident. The latter areas are considered accessible on a controlled exposure basis. Direct shine from the containment is less than 3.87E-06 c/kgh in the control room, technical support center, and counting facility and less than 1.29E-04 c/kgh in other vital areas in the Reactor Building.

(4) Access to radwaste is not required, but the Radwaste Building is accessible, since primary containment sump discharges are isolated and secondary containment sump pump power is shed at the onset of the accident. Thus, fission products are not transported to radwaste. The combustible gas control system is operated from the control room; the ABWR does not have a containment isolation reset control area or a manual ECCS alignment area. These functions are provided in the control room.

Area Gamma (Gy) Beta (Gy)

Primary Containment 2x106 2x107

ECCS Rooms 4x105 8x107

SGTS Area 5x105 2x10-1


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(5) Following an accident, access is available to electrical equipment rooms containing motor control centers and corridors in the upper Reactor Building (Subsection 12.3.6). This is based on radiation shine from the ECCS rooms and primary containment. While not necessary to maintain safe shutdown, such access can be useful in extending system functionality and plant recovery.

(6) The emergency power supplies (diesel generators) are accessible. However, access is not necessary, since the equipment is environmentally qualified.

1AA.3 Containment Description and Post Accident Operations

1AA.3.1 Description of Primary/Secondary Containment

The ABWR design includes many features to assure that personnel occupancy is not unduly limited and safety equipment is not degraded by post-accident radiation fields. These features are detailed in Tier 2 and only a brief summary description and Tier 2 reference are provided here for emphasis.

The configuration of the pressure suppression primary containment with the suppression pool maximizes the scrubbing action of fission products by the suppression pool. The particulate and halogen content of the primary containment atmosphere following an accident is thereby substantially reduced compared to the Regulatory Guide 1.3 source terms.

Primary containment leakage is limited to less than one half percent of the primary containment atmosphere per day. The surrounding secondary containment is kept at a negative pressure with respect to the environment permitting monitoring and treating all radioactive leakage from the primary containment.

The Standby Gas Treatment System (SGTS) operates automatically from the beginning of an accident to control the secondary containment pressure to -6.4 mm w.g. The large volume of this portion of the Reactor Building acts as a mixing chamber to dilute any primary containment leakage before processing by the SGTS and discharge to the environment. Discharge of radioactivity is thus controlled and reduced. Radioactivity content of secondary containment atmosphere is reduced with time by SGTS treatment as well as by decay. (However, prior removal of halogens by scrubbing in the suppression pool offsets the necessity of this treatment.)

Each ECCS pump and supporting equipment is located in an individual shielded, watertight compartment. Spread of radioactivity among compartments is thus limited. Radiation to the other equipment areas and corridors of the Reactor Building is limited to shine through the walls; there is no airborne radiation in these other areas. As these become accessible after an accident, any component failures can be repaired, thereby improving systems availability.


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1AA.3.2 Vital Area and Systems

A vital area is any area which will or may require occupancy to permit an operator to aid in the mitigation of or recovery from an accident. Areas which must be considered as vital after an accident are the control room, technical support center, sampling station, sample analysis area and the HPIN nitrogen supply bottles.

The vital areas also include consideration (in accordance with NUREG- 0737, II.B.2) of the post-LOCA hydrogen control system, containment isolation reset control area, manual ECCS alignment area, motor control center and radwaste control panels. However, the ABWR design does not require a containment isolation reset control area or a manual ECCS alignment area, as these functions are available from the control room. Those vital areas which are normally areas of mild environment, allowing unlimited access, are not reviewed for access.

Essential systems specific to the ABWR to be considered post-accident are those for the ECCS, fission product and combustible gas control and the auxiliary systems necessary for their operation (i.e., instrumentation, control and monitoring, power, cooling water, and air cooling).

1AA.3.3 Post Accident Operation

Post-accident operations are those necessary to (1) maintain the reactor in a safe shutdown condition, (2) maintain adequate core cooling, (3) assure containment integrity, and (4) control radioactive releases within 10CFR100 guidelines.

Many of the safety-related systems are required for reactor protection or to achieve a safe shutdown condition. However, they are not necessarily needed once a safe shutdown condition is achieved. Thus, the systems considered herein are the engineered safety features (ESF) (Chapter 6) used to maintain the plant in a safe shutdown condition.

For purposes of this review, the plant is assumed to remain in the safe shutdown condition.

The basis for this position is that the foundation of plant safety is the provision of sufficient redundancy of systems and logic to assure that the plant is shut down and that adequate core cooling is maintained. Necessary shutdown and post-accident operations are performed from the control room, except for the post-accident sampling station, the sample analysis area, and two manual nitrogen reserve supply valves.


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1AA.4 Design Review Bases

1AA.4.1 Radioactive Source Term and Dose Rates

The radioactive source term used is equivalent to the source terms recommended in Regulatory Guides 1.3 and 1.7 and Standard Review Plan 15.6.5 with appropriate decay times. Depressurized coolant is assumed to contain no noble gas. There is no leakage outside of secondary containment other than via the SGTS.

Dose rates for areas requiring continuous occupancy may be averaged over 30 days to achieve the desired <0.15 mSv/h.

Design dose rates for personnel in a vital area are such that the guidelines of General Design Criteria (GDC) 19 (i.e., <0.05 Sv whole body or its equivalent to any part of the body) are not exceeded for the duration of the accident, based upon expected occupancy and protection.

1AA.4.2 Accidents Used as the Basis for the Specified Radioactivity Release

Table 15.0-3 summarizes the various design basis accidents and associated potential for fuel rod failure. This chapter also provides the accident parameters. Of those accidents, only the DBA-LOCA may produce 100% failed fuel rods under NRC worst-case assumptions. The rod drop accident and fuel handling accident are the only other accidents postulated as leading to failed fuel rods with the potential consequence of radioactivity releases.

For the fuel handling accident, the reactor is either shutdown and cooled or is operating normally if the accident is in the spent fuel storage pool. Based on the conditions of Regulatory Guide 1.25, it is assumed that the airborne activity of the reactor building (Table 15.7-9) is released to the environment over a 2-hour period via a 99% iodine efficient SGTS. The total activity released to the environment is presented in Table 15.7-10 and the calculated exposure in Table 15.7-11. The exposures are within the guidelines of 10CFR100. Thus, recovery is possible well within the specified 100-day equipment qualification period. ECCS equipment is not affected by this accident and radiation in the ECCS area is not increased. This accident is not considered further.

The postulated control rod drop accident (Subsection 15.4.9) is one which occurs without a pipe break and so may require depressurization to attain long-term core cooling with the RHR System. Normally, this accident is terminated by a scram, and the plant is cooled and recovers. The performance of the separation-detection devices and the rod block interlocks virtually preclude the cause of a rod drop accident. This accident is not further considered.

The DBA-LOCA is the accident producing the limiting conditions of interest for this design review. In this accident the reactor is depressurized and reactor water mixes with


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suppression pool water in the process of keeping the fuel covered and cooled. Fission products are assumed to be essentially instantaneously released and mixed in the containment atmospheres and suppression pool-reactor water volumes.

1AA.4.3 Availability of Offsite Power

The availability of offsite power is not influenced by plant accident conditions. Loss of offsite power may be assumed as occurring coincident with the beginning of the accident sequence; however, continued absence of offsite power for the accident duration is not realistic. While restoration of offsite power is not a necessary condition for maintaining core cooling, its availability can permit operation of other plant systems which would not otherwise be permitted by emergency power restrictions (e.g., operation of the pneumatic air system, non-safety-related HVAC systems and other systems useful to plant cleanup and recovery).

Based on Table 19D.3-3, the probability for offsite power recovery is estimated to be very high in 8 hours. This is conservative, since the longest time for restoration of offsite power was six hours for the Pennsylvania-New Jersey-Maryland interconnection. The grid used as a basis for the probabilistic risk assessment is presented in Subsection 15D.3.

1AA.4.4 Radiation Qualification Conditions

The safety-related equipment requiring review for qualification is only that necessary for post-accident operations and for providing information for assuring post-accident control.

In 10CFR50, the long-term cooling capability is given as follows: “...decay heat shall be removed for the extended period of time required by the long lived radioactivity remaining in the core.” A 100-day period has been selected as a sufficient extended period permitting site and facility response to terminate the event.

As part of the design review process, a set of reference conditions is necessary for comparing expected post-accident radiation exposures. Appendix 3I defines the environmental conditions for safety-related equipment zones for periods of 60 years normal operations, including anticipated tests and abnormal events, and six months following the DBA-LOCA. These conditions are upper bound envelopes used to establish the environmental design and qualification bases of safety-related equipment. In effect, these are specification values, and equipment will be qualified to meet or exceed these values.

Radiation sources in the secondary containment (especially the ECCS rooms of the Reactor Building) are the same as the Table 1AA-1 design basis values for water sources. For airborne radiation sources, the plant design basis of Table 1AA-1 for air is used. Primary containment leakage is assumed to occur in each of the individual secondary containment compartments. This leakage is limited by the fission product control


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systems (Subsection 6.5.3). As previously noted, no credit has been taken for the radio-halogen scrubbing, which is an inherent feature of the BWR.

1AA.5 Results of the Review

1AA.5.1 Systems Required Post-Accident

This section establishes the various systems equipment which are required to function following an accident along with their locations. The expected habitability conditions and access and control needs are identified for the required post-accident period.

1AA.5.1.1 Necessary Post-Accident Functions and Systems

Following an accident and assuming that immediate plant recovery is not possible, the following functions*are necessary:

(1) Reactivity control

(2) Reactor core cooling

(3) Reactor coolant system integrity

(4) Primary reactor containment integrity

(5) Radioactive effluent control

Reactivity control is a short-term function and is achieved when the reactor is shutdown. The remaining functions are achieved in the longer term post-accident period by use of:

(a) The Emergency Core Cooling System (ECCS) and their auxiliaries (for reactor core cooling)

(b) The Combustible Gas Control System (CGCS) and auxiliaries (for primary containment and reactor coolant system integrity)

(c) The fission product removal and control system and auxiliaries (for radioactive effluent control)

(d) Instrumentation and controls and power for accident monitoring and functioning of the necessary systems and associated habitability systems

* ANSI/ANS 4.5 Criteria for Accident Monitoring Functions in Light Water Reactors.


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Tables 1AA-2 through 1AA-5 are generated to show:

(i) What major equipment and systems are required to function and thereby define the systems for review

(ii) The redundant equipment locations by divisional isolated room or area and containment or building

1AA.5.1.2 Emergency Core Cooling Systems and Auxiliaries

Table 1AA-2 lists various systems related to cooling the fuel under post-accident conditions as described in Section 6.3 and Subsection 9.4.5.2 (HVAC). This table shows ECCS equipment and equipment coolers in an ECCS room. Instrumentation transmitters are in adjoining areas. The required power and cooling water in the same division are described in Subsection 1AA.5.1.5. All perform together to provide an ECCS function.

The Automatic Depressurization System (ADS) function is described in Subsection 1.2.2.2.2.4. A postulated non-break or small break accident could require continued need for the depressurization function until the RHR System is placed in the shutdown reactor cooling mode. In the case of a non-break or a small break accident, the majority of the fission products would be released via the safety/relief valves to the suppression pool and hence to the containment, rather than direct mixing through the supersession pool vents, as would occur following a DBA-LOCA. In either case, the distribution of fission products is assumed to be the same as for the DBA-LOCA even though, realistically, a significant portion of halogens and solid fission products would be retained in the reactor pressure vessel. Thus, the results as they apply to the ADS are very conservative. The pneumatic nitrogen supply for the ADS and other containment valves is included in Table 1AA-3 as a portion of the combustible gas control. The hand-operated nitrogen reserve supply valves P54-F017C and D are accessible outside the secondary containment, if needed, to mitigate a large leak.

The high pressure core flooder (HPCF) and the low pressure flooder loop (LPFL) functions are described in Subsections 1.2.2.5.2 and 1.2.2.5.1.1, respectively. The cooling function can also satisfy the containment cooling function in that, by cooling suppression pool water, which is the source of water flowing to the reactor, the containment source of heat is also removed. The wetwell/drywell sprays are described in Subsection 1.2.2.5.1.3.

The fuel pool cooling function (Subsection 1.2.2.7.2) is also included on the basis that a recently unloaded fuel batch could require continued cooling during the post-accident period. The equipment is environmentally qualified, so access is not required and redundancy is included in system components.

The locations of selected associated valves and instrument transmitters are included. These do not represent all of this type of equipment which is environmentally qualified,


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safety-related, or included in the systems of Table 3.2-1. It does however, represent principal components which are needed to operate, generally during post accident operations. For example, most ECCS valves are normally open, and only a pump discharge valve needs to open to direct water to the reactor. Similarly, the instrument transmitters shown are those which would provide information on long-term system performance post-accident. Control room instrumentation is not listed, since it is all in an accessible area where no irradiation degradation would be expected. Passive elements such as thermocouples and flow sensors are not listed although they are environmentally qualified. The components listed under main steam (B21) are those for ECCS function or monitoring reactor vessel level. Suppression pool level is included with the HPCF instrumentation.

1AA.5.1.3 Combustible Gas Control Systems and Auxiliaries

Flammability control in the primary containment is achieved by an inert atmosphere during all plant operating modes except operator access for refueling and maintenance and a recombiner system to control oxygen produced by radiolysis. The high pressure nitrogen (HPIN) gas supply is described in Subsection 1.2.2.12.13. The Containment Atmospheric Monitoring System (CAMS) measures and records containment oxygen/hydrogen concentrations under post-accident conditions. It is automatically initiated by detection of a LOCA (Subsection 7.6.1.6). Table 1AA-3 lists the combustible gas control principal components and their locations.

1AA.5.1.4 Fission Product Removal and Control Systems and Auxiliaries

Engineered Safety Feature (ESF) filter systems are the Standby Gas Treatment System (SGTS) and the control building Outdoor Air Cleanup System. Both consist of redundant systems designed for accident conditions and are controlled from the control room. The SGTS filters the gaseous effluent from the primary and secondary containment when required to limit the discharge of radioactivity to the environment. The system function is described in Subsection 1.2.2.15.4.

A portion of the Control Building heating ventilating and air-conditioning (HVAC) provides detection and limits the introduction of radioactive material and smoke into the control room. This portion is described Subsection 9.4.1.1.3.

The CAMS described in the previous section also measures and records containment area radiation under post-accident conditions. A post-accident sampling system (PASS) obtains containment atmosphere and reactor water samples for chemical and radiochemical analysis in the laboratory. Delayed sampling, shielding, remote operated valves and sample transporting casks are utilized to reduce radiation exposure. The samples are manually transported between the PASS room in the Reactor Building and the analysis laboratory in the service building. The system is described in


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Subsection 9.3.2.3.1. Table 1AA-4 lists the fission product removal control components and locations.

1AA.5.1.5 Instrumentation and Control, Power, and Habitability Systems and Auxiliaries

Most of the post-accident instrumentation and control system equipment is listed with the applicable equipment in Tables 1AA-2, 1AA-3 and 1AA-4. The remaining instrumentation and control equipment is included with the power and habitability systems equipment listed in Table 1AA-5. Instrumentation is consistent with the post accident phase variables monitored by the Post-Accident Monitoring (PAM) System listed in Table 7.5.2.

Standby AC power is supplied by three diesel generators in separate electrical divisions (Subsection 1.2.2.13.13). The diesel generators, switchgear and motor control centers are included in the unit Class 1E AC power system described in Subsection 1.2.2.13.14.1. Storage batteries are the standby power source for the unit Class 1E DC power system described in Subsection 1.2.2.13.12.2. The safety system logic and control power system is described in Subsection 1.2.2.13.14.1.

Habitability systems ensure that the operator can remain in the control room and take appropriate action for post-accident operations. The control building includes all the instrumentation and controls necessary for operating the systems required under post- accident conditions.

The control room, control and reactor building HVAC essential equipment are a portion of the plant environmental control of temperature, pressure, humidity and airborne contamination described in Subsection 1.2.2.16.5 (1), (4), (5), (7) and (8). HVAC units controlling the local room environments are included with respective equipment in Tables 1AA-2, 1AA-3 and 1AA-4. The major HVAC equipment and locations are listed in Table 1AA-5.

The Reactor Building Cooling Water (RBCW) System provides cooling water to designated equipment in the Reactor Building (including containment) as described in Subsection 1.2.2.12.3. The HVAC Emergency Cooling Water (HECW) System provides chilled water to designated equipment in the control building as described in Subsection 1.2.2.12.6.


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* Uniformly mixed within the primary containment boundary

† Uniformly mixed in the suppression pool and reactor coolant

Table 1AA-1 Radiation Source Comparison

Activity Group % Core Inventory Released

R.G.1.3 R.G.1.7 Plant Design Basis

Air

Noble Gases 100 100 100*

Halogens 25 — 25*

All Remaining — — —

Water

Noble Gases 0 — 0

Halogens — 50 50†

All Remaining — 1 1†


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Table 1AA-2 Post-Accident Emergency Core Cooling Systems and

Auxiliaries

Equipment MPL Location

ADS & Transmitters

SR Valve B21-F010A,C,F,H,L,N,R,T Upper Drywell (PC)

SR Accumulator B21-A004A,C,F,H,L,N,R,T Upper Drywell (PC)

Rx Water Level (ADS,RHR) B21-LT003A thru H Instrument Rack Rm. (SC)

Rx Water Level (HPCF) B21-LT001A,B,C,D Instrument Rack Rm. (SC)

Rx Pressure (RHR) B21-PT301A,B,C,D Instrument Rack Rm. (SC)

DW Pressure (HPCF, RHR) B21-PT025A,B,C,D Instrument Rack Rm. (SC)

HPCF

Pumps E22-C001B,C HPCF Rm. B,C (SC)

SP Suction Valve E22-F006B,C HPCF Rm. B,C (SC)

Rx Injection Valve E22-F003B,C Valve Rm. B,C (SC)

CST Suction Valve E22-F001B,C Valve Rm. B,C (SC)

Essential HVH (HVAC) U41-D102,106 HPCF Rm. B,C (SC)

CST Water Level P13-LT001A,B,C,D HPCF Rm. B,C (SC)

Flow E22-FT008B1,B2,C1,C2 By HPCF Rm. B,C (SC)

Suction Pressure E22-PT002,003; B,C By HPCF Rm. B,C (SC)

Injection Pressure E22-PT006,007; B,C By HPCF Rm. B,C (SC)

LPCF

Pump E11-C001A,B,C RHR Rm. A,B,C (SC)

Heat Exchanger E11-B001A,B,C RHR Rm. A,B,C (SC)

RCW Discharge Valve P21-F013A,B,C RHR Rm. A,B,C (SC)

SP Suction Valve E11-F001A,B,C RHR Rm. A,B,C (SC)

Rx Injection Valve E11-F005A,B,C Valve Rm. A,B,C (SC)

Rx Return Valve E11-F010,011,012;A,B,C Valve Rm. A,B,C (SC)

DW Spray Valve E11-F017,018;B,C Valve Rm. B,C (SC)

WW Spray Valve E11-F019B,C Valve Rm. B,C (SC)

FPC Supply Valve E11-F015B,C Valve Rm. B,C (SC)

FPS Supply Valve E11-F101,102,103 Valve Rm. B,C (SC)

Essential HVH (HVAC) U41-D103,104,105 RHR Rm. (SC)

Flow E11-FT008A1,B1,C1 By RHR Rm. A,B,C (SC)

Flow E11-FT008A2,B2,C2 By RHR Rm. A,B,C (SC)


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(PC)—Primary Containment

(SC)—Secondary Containment

RCW Flow P21-FT008A,B,C By RHR Rm. A,B,C (SC)

Hx I/O Temperature E11-TT006,007;A,B,C By RHR Rm. A,B,C (SC)

Discharge Pressure E11-PT004A thru G By RHR Rm. A,B,C (SC)

DW Temperature T31-TT/SSA051,053 Inst. Rack Rm. (SC)

DW/WW Pressure Ratio T31-PT055A,B Inst. Rack Rm. (SC)

WW Pressure T31-PT056A,B Inst. Rack Rm. (SC)

DW Pressure T31-PT054 Inst. Rack Rm. (SC)

FPCS

Pump G41-C001A,B FPC Pump Rm. (SC)

Heat Exchanger G41-B001A,B FPC Hx Rm. (SC)

Pump Discharge Valve G41-F021A,B FPC Valve Rm. (SC)

Essential HVH (HVAC) U41-D107, 108 FPC Valve Rm. (SC)

Flow G41-FT006A,B By FPC Pump Rm. (SC)

Suction Pressure G41-PT003A,B By FPC Pump Rm. (SC)

Skimmer ST Level G41-LT020 Refueling Floor(SC)

Table 1AA-2 Post-Accident Emergency Core Cooling Systems and

Auxiliaries (Continued)



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(PC)—Primary Containment


(RB)—Reactor Building outside (Secondary Containment)

Table 1AA-3 Post-Accident Combustible Gas Control Systems and Auxiliaries


HPIN

Nitrogen Storage Bottles P54-A001A Thru V By Valve Rm (RB)

Supply Pressure P54-PT002A, B, 004, 005 By Valve Rm (RB)

FCS

Recombiner & Auxiliaries T49-A001A,B (PC)

RHR Cooling/Isol. Valve T49-F008,010; A,B (PC)(SC)

Flow T49-FT002,004; A,B Inst. Rack Rm. A,B (SC)

Pressure T49-PT003A,B Inst. Rack Rm. A,B (SC)

CAMS

Hydrogen, Oxygen Elements D23-H2, O2 Rack A,B CAMS Rm. A,B (SC)

Gas Measurement D23-Gas Cal. Rack A,B CAMS Rm. A,B (SC)

Gas Elements D23-Gas Cal. Rack A,B CAMS Rm. A,B (SC)

DW Gas Valve D23-F004A,B CAMS Rm. A,B (SC)

WW Gas Valve D23-F006A,B CAMS Rm. A,B (SC)

Essential HVH (HVAC) U41-D113,114 CAMS Rm. A,B (SC)

Gas Supply D23-Gas Cyl. Rack A,B CAMS Rm. A,B (RB)


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(CB)—Control Building



(SB)—Service Building

Table 1AA-4 Post-Accident Fission Product Removal and Control Systems

and Auxiliaries


SGTS

Exhaust Fan T22-C001B, C Fan/Dryer Rm. (SC)

Charcoal Filter T22-D002C,D001B Filter Train Rm. (SC)

PC Inlet Valve T22-F002A,B Fan/Dryer Rm. (SC)

SC Inlet Valve T22-F001A,B Fan/Dryer Rm. (SC)

Stack Outlet Valve T22-F004A,B Filter Train Rm. (SC)

PC (DW,WW) Isolation Valves T31-F004,006,008 Valve Rm. (SC)

Essential HVH (HVAC) U41-D111,112 SGTS HVH Rm. (SC)

Radiation (Ion/Scint.) D11-RE002,011;A,B SGTS Monitor Rm. (RB)

Sampling Rack H22-P250 By SGTS (SC)

Flow T22-FT018B,C By Filter Train. Rm. (RB)

Filter Moisture T22-MT011B,C & 012B,C By Filter Train. Rm. (RB)

CR HVAC

Emerg. Recirculation Fan U41-C603B CR HVAC Rm. A,B (CB)

Emerg. Charcoal Filter Unit U41-B A,B CR HVAC Rm. A,B (CB)

Air Intake Isolation Valves U41-F A,B CR HVAC Rm. A,B (CB)

PASS

Conditioning/Holding Rack P91 (SC)

Sampling/Casks Rack P91 PASS Rack Rm. (RB)

LPCF Supply Valve E11-F045,046; A (SC)

DW/WW Gas (CAMS) Valve D23 (SC)

Control Panel (PT,TT) H22 PASS Rack Rm. (RB)

Chemical Radiological Analysis Laboratory (SB)

Stack

Radiation (Ion/Scint.) D11-RE041,043; A,B Stack (RB)

Monitor Racks, Control Rod H21,H22 Stack Monitoring Rm.(RB)


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Table 1AA-5 Post-Accident Instrumentation and Controls, Power and Habitability

Systems and Auxiliaries


Instrumentation & Controls

Post-Accident I&C H11-Post-Accident Control & Panel Rms. (CB)

Power

DC Supply R42-Storage Batteries Battery Rm. (CB)

ESF HV&LV Switchgear R22-Post-Accident Emerg. Electric Rm. A,B,C (RB)

ESF Motor Control Center R24-Post-Accident Emerg. Electric Rm. A,B,C (RB)

Diesel Generator & Auxiliaries R43-DG A,B,C DG Rm. A,B,C (RB)

DG Motor Control Center R43-P001A,B,C DG MCC Rm. A,B,C (RB)

Supply Fan (HVAC) U41-C201A,E,204B,F 207C, G

DG Supply Fan Rm. A,B,C (RB)

Exhaust Fan (HVAC) U41-202A,E,205B,F 207C, G

DG Exhaust Fan Rm. A,B,C (RB)

Essen. Fresh Air Fan (HVAC) U41-203A,E,206B,F 209C,G

DG Essen. Fan Rm. A,B,C (RB)

RCW Discharge Valve P21-F055A thru F DG Rm. A,B,C (RB)

Control Panel (H21) R43-P002,003C,004;A,B,C DG Control Pnl. Rm. A,B,C (RB)

CB HVAC

Supply Fan U41-C606B,F,608C,G E/HVAC Rm. A,B,C (CB)

Exhaust Fan U41-C607B,F,609C,G E/HVAC Rm. A,B,C (CB)

MCR Supply Fan U41-C601B,F,604A,E CR HVAC Rm. A,B (CB)

MCR Exhaust Fan U41-C602B,F,605A,E CR HVAC Rm. A,B (CB)

RCW

Pump P21-C001A thru G Pump Rm. A,B,C (CB)

Hx Return Valve P21-F004D, E, F, A, B, C, D, G, H, J

Hx Rm. A,B,C (CB)

Non-Post-Accident Supply Valve P21-F074A,B,C (RB)

Non-Post-Accident Return Valve P21-F082A,B,C (RB)

Flow P21-FT006A,B,C By Pump Rm. A,B,C (CB)

Pressure P21-PT004A,B,C By Pump Rm. A,B,C (CB)

Surge Tank Level P21-LT013A,B,C By Surge Tank A,B,C (RB)

HECW

Pump P25-C001A,B,C,E,F Chiller Rm. A,B,C (CB)


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(CB)—Control Building

Refrigerator P25-D001A,B,C,E,F Chiller Rm. A,B,C (CB)

Pressure Control Valve P25-F012 B,C HVAC Rm. A,B,C (CB)

Temperature Control Valve P25-F005 B,C HVAC Rm. A,B,C (CB)

Temperature Control Valve P25-F016 A,B,C HVAC Rm. A,B,C (CB)

Temperature Control Valve P25-F022 A,B,C (RB)

RCW Temp. Control Valve P21-F025 A,B,C,E,F (CB)

Instrument Air

Compressor P52-C001,002 Inst.Air Rm. (RB)

Table 1AA-5 Post-Accident Instrumentation and Controls, Power and Habitability

Systems and Auxiliaries (Continued)


/18

Not Used 1B-1

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1B Not Used

/2

ABWR Station Blackout Considerations 1C-1

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1C ABWR Station Blackout Considerations

1C.1 Introduction

This appendix describes (a) how the ABWR Design addresses Station Blackout (SBO) Events; (b) how the ABWR Design complies with 10CFR50.63 SBO requirements; and (c) where supporting documentation to these conformances exist in Tier 2.

1C.2 Discussion

1C.2.1 Station Blackout (SBO) Definition

For the ABWR design the definitions of Station Blackout, Alternate AC (AAC) Power Source, and Safe Shutdown given in 10CFR50.02 are provided below:

■ Station Blackout

“Station blackout means the complete loss of alternating current (AC) electric power to the essential and nonessential switchgear buses in a nuclear power plant (i.e., the loss of offsite electric power system concurrent with turbine trip and unavailability of the onsite emergency AC power system). Station blackout does not include the loss of available AC power to buses fed by station batteries through inverters or by alternate AC sources as defined in this section, nor does it assume a concurrent single failure or design basis accident.”

■ Alternate AC Power Source

“Alternate AC source means an alternating current (AC) power source that is available to and located at or nearby a nuclear power plant and meets the following requirements:

(1) Is connectable to but not normally connected to the offsite or onsite emergency AC power systems

(2) Has minimum potential for common mode failure with offsite power or the onsite emergency AC power sources

(3) Is available in a timely manner after the onset of station blackout

(4) Has sufficient capacity and reliability for operation of all systems required for coping with station blackout and for the time required to bring and maintain the plant in safe shutdown (non-design basis accident)”

1C-2 ABWR Station Blackout Considerations

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■ Safe Shutdown (SSD)

“Safe shutdown (non-design basis accident (non-DBA)) for station blackout means bringing the plant to those shutdown conditions specified in plant technical specifications as Hot Standby or Hot Shutdown, as appropriate…”

1C.2.2 Plant SBO Design Basis

1C.2.2.1 General SBO Design Basis

■ The ABWR design will mitigate station blackout events as defined in Subsection 1C.2.2.

■ The ABWR design will comply with 10CFR50.63 requirements relative to the loss of all alternating current power sources.

■ The ABWR design will include and utilize an Alternate AC (AAC) power source to comply with 10CFR50.63 requirements and the recommendations for ALWRs, as defined by the NRC in SECY 90-016.

■ The ABWR design will be consistent with Regulatory Guide 1.155 and NUMARC 87-00 guidelines relative to an AAC power source.

■ The ABWR design AAC power source will supplement and compliment the current offsite AC power connections, the onsite normal AC power sources (the unit auxiliary and reserve auxiliary transformers), the onsite emergency AC power sources (DGs) and the onsite DC power sources.

1C.2.2.2 Specific SBO Design Basis

■ The ABWR AAC power source will be a combustion turbine generator (CTG).

■ The normal design function of the CTG will be to act as a standby, non-safety-related power source for the plant investment protection (PIP) non-safety-related loads during loss of preferred power (LOPP) events.

■ The CTG will be capable of being manually configured to provide power to a selected safety-related emergency bus within 10 minutes during SBO events.

■ The CTG will automatically start, accelerate to required speed, reach required voltage and frequency and be ready to accept PIP loads within two minutes of the receipt of its start signal.

■ The CTG will be a diverse, self contained unit (including its auxiliaries) and will be independent of the plant preferred and emergency power sources.


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■ The target reliability of the GTG will be >0.95, as calculated by NSAC-108 methodology.

■ The CTG will have capacity to supply the required safe shutdown loads.

■ The CTG will be housed in a Uniform Building Code structure which is protected from adverse site weather related conditions.

■ The CTG design will minimize potential for single point failure vulnerability with onsite emergency power sources.

■ Adequate pneumatic pressure and water makeup sources will be available throughout the SBO duration.

■ The ABWR design will confine the SBO duration to 10 minutes or less with the use of the AAC power source.

■ The CTG will be controllable locally or from the MCR.

■ Provisions will be made to facilitate the orderly restoration of offsite and onsite power source during the SBO event.

■ Special quality assurance and control practices will be applied to the CTG.

■ Special equipment requirements will be applied to the CTG support components.

■ The CTG will utilize a separate fuel oil storage tank and transfer system from that of the onsite emergency power sources.

■ The CTG will operate during the SBO event without external AC power sources.

■ The standby function of the CTG will be to mitigate LOPP or SBO events.

■ Dual manually operated circuit breakers will separate the CTG from the onsite emergency power buses.

■ The AAC power source will utilize the available station and/or internal batteries for breaker control and initial CTG starting functions.

■ The CTG Fuel Oil Supply will be periodically inspected and the oil analyzed.

■ The CTG operation will be subject to plant operation, maintenance and testing procedures.

■ All operator actions required during SBO events will be demonstrated by training exercises and will be according to appropriate plant procedures.


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■ CTG power will be used to restore various selected plant environmental control components (HVAC, chillers, etc.) as soon as possible.

■ The CTG will not normally be used to provide power connected to the plant loads.

■ The CTG will be capable of being inspected, tested, and maintained.

■ The CTG capabilities will be demonstrated prior to shipment, during initial pre-operational test, and periodically during power operation.

■ Required plant core cooling and containment integrity during the SBO duration (10 minutes) will not depend on any AC power sources.

1C.2.3 Plant SBO Safety Analysis

1C.2.3.1 Plant Event Evaluations

1C.2.3.1.1 Plant Normal Operation

The normal configuration of the onsite AC power distribution system and its individual power sources are described in Subsections 8.2.1 and 8.3.1. The CTG (AAC) system attributes and its interconnections are described in Subsection 9.5.11 and in Subsection 8.3.1, respectively. Both are shown on Figure 8.3-1.

The normal and alternate preferred AC power sources supply safety-related and non-safety-related loads. Power to these loads are supplied from the unit auxiliary transformers (UATs) units and the reserve auxiliary transformer (RAT).

The CTG is designed to supply standby power to the non-Class 1E 6.9 kV buses which carry the plant investment protection (PIP) loads. The CTG automatically starts on detection of under voltage on the PIP buses. When the CTG is ready to assume load, if the voltage is still deficient, power automatically transfers to the CTG (refer to Figure 8.3-1).

The CTG can also supply standby power to the non-Class 1E 6.9 kV power generation buses which supply feedwater and condensate pumps. These buses normally receive power from the unit auxiliary transformers and supply power to the plant investment protection (PIP) buses through a cross-tie. The cross-tie automatically opens on loss of power but may be manually reclosed if it is desired to operate a condensate pump from the combustion turbine generator or the reserve auxiliary transformer which are connectable to the PIP buses. This arrangement allows the powering of load groups of non-Class 1E equipment in addition to the Class 1E divisions which may be used to supply water to the reactor vessel (refer to Figure 8.3-1).


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1C.2.3.1.2 LOPP Events

The ABWR onsite emergency power sources during LOPP events are the diesel generator (DG) units. These units and their system responses are discussed in Subsection 8.3.1.1.8. However, the CTG is available to provide backup emergency power during LOPP to safety-related loads by manual reconfiguration of the CTG and the loads.

1C.2.3.1.3 SBO Events

The CTG is the AC power source during an SBO event. The CTG can supply 6.9 kV Class 1E buses through the realignment of pre-selected breakers during SBO events. The CTG will reach operational speed and voltage in 2 minutes and will be available for bus connection within 10 minutes. Upon a LOPP, the CTG is automatically started and configured to non-safety-related PIP loads. Plant operators using appropriate procedures will reconfigure any of the 6.9 kV Class 1E buses to accept CTG power. Refer to Tier 2 Subsections 8.3.1.1.7 and 9.5.11.

1C.2.3.1.4 Other Operational Capabilities

The CTG can be used for postulated prolonged SBO scenarios.

Up to the limits of its capacity, the CTG can be connected to any combination of Class 1E and non-Class 1E buses to supply loads in excess of the minimum required for safe shutdown.

The ABWR design provides for local and main control room operation of the CTG. Communication is available between the CTG area and the main control room.

1C.2.3.2 Alternative AC Power Source Evaluation

The alternate AC power source (1) is a combustion turbine generator, (2) is provided with an immediate fuel supply that is separate from the fuel supply for other onsite emergency AC power systems, (3) fuel will be sampled and analyzed consistent with applicable standards, (4) is capable of operating during and after a station blackout without any AC support systems powered from the preferred power supply or the blacked-out units Class 1E power sources (5) is designed to power all of the PIP and/or Class 1E shutdown loads necessary within 10 minutes of the onset of the station blackout, such that the plant is capable of maintaining core cooling and containment integrity (6) will be protected from design basis weather events (except seismic and tornado missiles) to the extend that there will be no common mode failures between offsite preferred sources and the combustion turbine generator power source, (7) will be subject to quality assurance guidelines commensurate with its importance to SBO, (8) will have sufficient capacity and capability to supply one division of Class 1E loads, (9) will have sufficient capacity and capability to supply the required non-Class 1E loads


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used for a safe shutdown, (10) will undergo factory testing to demonstrate its ability to reliably start, accelerate to required speed and voltage and supply power within two minutes, (11) will not normally supply power to nuclear safety-related equipment except under specific conditions, (12) will not be a single point single failure detriment to onsite emergency AC power sources, and (13) will be subject to site acceptance testing; periodic preventative maintenance, inspection, testing; operational reliability assurance program goals.

Based on the above, the ABWR design for the AAC power supply complies with 10CFR50.63, with Regulatory Guide 1.155 and with NUMARC 87-00 and meets the SBO rule.

1C.2.4 Plant Conformance With SBO Requirements

A brief review of the general ABWR design conformance with various SBO requirements and guidelines is given below. A more complete in-depth and specific review of each of the SBO regulatory requirements or guidelines is given in the enclosed tables (refer to Tables 1C-1 through 1C-3).

1C.2.4.1 10CFR50.63 Requirements

The ABWR complies with the 10CFR50.63 requirements. Special attention was given to the regulation definition of the SBO event, the event conditions, and the requirement for safe shutdown status. The ABWR utilizes the AAC power source option and provides an evaluation of the requirements/compliances in Table 1C-1.

1C.2.4.2 New ALWR Requirements (SECY 90-016)

A review of the new ALWR SBO requirements in SECY 90-016 recommendations was conducted. The ABWR design is in compliance with the ALWR recommendations.

1C.2.4.3 Regulatory Guide 1.155 Guideline Requirements

A review of the ABWR CTG design relative to Sections 3.3.5, 3.3.6, 3.3.7, 3.4 and Appendix A and B of RG 1.155 was conducted. CTG design fully complies with the cited requirements. The use of the CTG as an AAC power source in the ABWR design eliminates the need for a SBO coping analyses by limiting the SBO duration to 10 minutes or less. No operator action is required within the initial ten minutes (refer to Table 1C-2).

1C.2.4.4 NUMARC 87-00 Guidelines

A review of the ABWR CTG design relative to the NUMARC SBO guidelines, Subsections 7.1.1 and 7.1.2 and Appendices A and B was conducted. The ABWR design with CTG is consistent with the NUMARC guidelines (refer to Table 1C-3).


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1C.2.5 Other SBO Considerations

Several other SBO considerations are identified below for special compliance or consideration.

1C.2.5.1 Plant Technical Specifications

Surveillance and operational requirements are needed for the CTG in order to assure its reliability or maintainability. However, these will be part of the COL applicant maintenance, testing, and inspection procedures. These procedures will not be part of technical specifications.

1C.2.5.2 Design Interface Requirements

The CTG has a limited number of design interface requirements. Fuel oil is initially supplied from a local tank, and then transferred from a fuel oil storage tank, both of which are independent of the DG fuel oil tanks. A seven (7) day oil supply for the CTG sufficient for shutdown loads will be available onsite. The local CTG I&C is powered by the unit itself or supplied from station batteries. Other auxiliary functions are an integral part of the CTG unit.

1C.2.5.3 Station Blackout Procedures

Appropriate procedures will include the use of the CTG and are the COL applicant’s responsibility. The procedures will consider specific instructions for operation actions responses, timing and related matters during SBO events. The operator actions will include power source switching, load shedding, etc. See Subsection 1C.4.1 for COL license information requirements.

1C.2.5.4 Equipment Qualification, Testing and Reliability

The CTG will be qualified (as a non-Class 1E AAC power source) for its intended duties and service. Qualification testing, equipment inspections, and reliability data will be made available.

1C.2.5.5 Periodic Surveillance, Testing, Inspection and Maintenance

Operational reliability assurance program (ORAP) requirements will be established for the CTG.

1C.2.5.6 Power and Control Cable Routing

The CTG power and control cable routing is physically and electrically separated from other power sources to the extent practical. A suggested routing is shown in Figure 8.2-1.


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1C.2.5.7 Plant Battery Recharging

The CTG is capable of recharging the plant batteries during SBO scenarios while supplying safe shutdown loads.

1C.2.5.8 Plant HVAC Restoration Capabilities

The CTG is capable of restoring environmental control components during the SBO duration while supplying the safe shutdown loads.

The Main Control Room environment will not exceed its design basis temperature even during a prolonged SBO event. With the CTG available in ten minutes, MCR HVAC can be restored.

1C.2.5.9 Circuit Breaker Operation

During the realignment of the CTG from non-safety-related buses to safety-related buses, at least two breakers will need to be manually closed. One of these breakers is Class 1E, and is controlled by the Class 1E battery power within the same division. The other breaker is non-Class 1E, and is controlled by the non-Class 1E battery.

The current SBO requirement that at least one emergency bus be powered within ten minutes is achieved by the manual operation of the two breakers between the CTG and the selected emergency bus (see Figure 8.3-1).

In order to maintain a minimum number of direct connections between the CTG and any of the three Class 1E emergency buses, only one Class 1E bus has its supply breaker racked in. It can therefore be controlled directly from the main control room. The other emergency buses have their supply breakers racked out, and therefore, require local operator action to rack in the breakers before main control room operation is available.

1C.2.5.10 CTG – Physical Protection Considerations

The CTG is housed in a building (separate from the building which contains the DGs) above the design flood levels. The building is designed to protect the CTG from site related weather conditions.

1C.3 Conclusions

In summary:

■ The ABWR design will utilize a combustion turbine generator (CTG) as its Alternate AC (AAC) power source in complying with 10CFR50.63 SBO.

■ The ABWR design complies with 10CFR50.63 and RG 1.155 and is consistent with NUMARC 87-000 guidelines.


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■ The ABWR design can successfully prevent or mitigate the consequences of an SBO event.

1C.4 COL License Information

1C.4.1 Station Blackout Procedures

The COL applicant shall provide procedures for SBO events including the use of the CTG as described in Subsection 1C.2.5.3.

1C.5 References

1C-1 SECY-90-016, “Evolutionary LWR Certification Issues and Their Relationship to Current Regulatory Requirements”, January 12, 1990.

1C-2 Letter J. Taylor to S. Chilk, “Evolutionary LWR Certification Issues and Their Relationship to Current Regulatory Requirements”, June 26, 1990.

1C-3 10CFR50.63, “Loss of All Alternating Current Power (Station Blackout-SBO)”, July 21, 1988.

1C-4 RG-1.155, “Station Blackout”, July 1988.

1C-5 NUMARC 87-00, “Guidelines and Technical Bases for NUMARC Initiative Addressing Station Blackout at LWRs” Plus Supplemental Questions and Answers, January 4, 1990.

1C-6 10CFR50.02, Definitions.

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Table 1C-1 ABWR Design Compliance with 10CFR50.63 Regulations

Requirements Compliance

§ 10CFR50-63 Loss of all alternating current power.

§ 50.63 Loss of all alternating current power.

(a) Requirements

(1) Each light-water-cooled nuclear power plant licensed to operate must be able to withstand for a specified duration and recover from a station blackout as defined in §

50.2. The specified station blackout duration shall be based on the following factors:

The ABWR design will utilize an alternate AC (AAC) power source to mitigate and recover from station blackout events (defined in 50.2). The AAC power source will be a combustion turbine generator (CTG). The CTG will be totally independent from offsite preferred and onsite Class 1E sources. A ten (10) minute interval is used as the ABWR design basis for the SBO event duration. The AAC power source provides a diverse power source to the plant.

(i) The redundancy of the onsite emergency AC power sources The ABWR design CTG will have sufficient capacity and capabilities to power the necessary reactor core coolant, control and protective systems including station battery and other auxiliary support loads needed to bring the plant to a safe and orderly shutdown condition (defined in 50.2). The CTG supplied will be rated at a minimum of 9 MWe and be capable of accepting shutdown loads within 10 minutes.

The current plant onsite emergency power sources include three (3) independent and redundant DG divisions which are designed to supply approximately 5 MWe within 1 minute.

Additionally, the plant has been designed to accommodate AC power source losses for a period up to 8 hours. The AAC limits the SBO event to 10 minutes.

(ii) The reliability of the onsite emergency AC power sources The current onsite emergency AC power sources will have the following reliability:

DGs…0.975

The CTG will have the following reliability:

CTG…0.95

The above values are used in the ABWR-PRA analysis.

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(iii) The expected frequency of loss of offsite power The expected frequency of loss of offsite power assumed was 0.1 events/yr.

(iv) The probable time needed to restore offsite power The offsite power is expected to be restored within 8 hours.

(2) The reactor core and associated coolant, control, and protection systems, including station batteries and any other necessary support systems, must provide sufficient capacity and capability to ensure that the core is cooled and appropriate containment integrity is maintained in the event of a station blackout for the specified duration. The capability for coping with a station blackout of specified duration shall be determined by an appropriate coping analysis. Utilities are expected to have the baseline assumptions, analyses, and related information used in their coping evaluations available for NRC review.

The AAC power source is capable of providing the necessary core, containment and equipment services (e.g. makeup and cooling water, I&C power, etc.) to bring the reactor to hot shutdown and then to cold shutdown conditions. The AAC will limit the SBO duration to 10 minutes.

The current plant design assures that during the 10-minute interval, the plant core, containment and other safety functions will be maintained without the use or need for AC power.

However, the AAC can operate indefinitely. A seven (7) day supply of oil sufficient for shutdown loads is available on site. Subsequent oil deliveries will be provided.

Table 1C-1 ABWR Design Compliance with 10CFR50.63 Regulations (Continued)


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(b) Limitation of scope

(c) Implementation

(1) Information Submittal. For each light-water-cooled nuclear power plant licensed to operate after the effective date of this amendment, the licensee shall submit the information defined below to the Director by 270

days after the date of license issuance.

In addition to the discussion under (a) above, the following is noted. The ABWR design SBO duration time considerations are consistent with RG1.155 and NUMARC-87-00. Upon loss of offsite power (LOPP) and upon the subsequent loss of all on site AC emergency power sources (three independent and redundant DGs), the CTG can be manually connected to any one of the three safety-related (Class 1E) busses by closing two circuit breakers. The alternative AC (AC) power source will automatically start, and within 2 minutes be up to required speed and voltage. It will then automatically connect to selected PIP buses (non-Class 1E) loads.

During the first 10 minutes, the reactor will have automatically tripped, the main steam isolation valves (MSIVs) closed, and the RCIC actuated.

The RCIC system will automatically control reactor coolant level. Any necessary relief valve operation will also be automatic.

Within the 10 minute SBO interval, none of the above actions will require AC power or manual operator actions.

The reconfiguration of the CTG to pick up the Class 1E buses will require manual closure of two circuit breakers from the control room. Upon restoration of power to the safety bus(es), the remaining safe shutdown loads will be energized.

(i) A proposed station blackout duration to be used in determining compliance with paragraph (a) of this section, including a justification for the selection based on the four factors identified in paragraph (a) of this section

(ii) A description of the procedures that will be implemented for station blackout events for the duration determined in paragraph (c)(1)(i) of this section and for recovery therefrom

Appropriate plant procedures will be developed by the COL applicant for the ABWR design. These procedures be integrated/coordinated with the plant EOPs, using the EOP methodology. Procedures will consider instructions for operator actions, responses, timing, and related matters during the SBO event.



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(iii) A list of modifications to equipment and associated procedures, if any, necessary to meet the requirements of paragraph (a) of this section, for the specified station blackout duration determined in paragraph (c)(1)(i) of this section, and a proposed schedule for implementing the stated modifications

Modifications to equipment and procedures is not applicable since the use of an AAC source and other SBO considerations are included in the ABWR design.

(2) Alternate AC source: The alternate AC power source(s), as defined in §

50.2, will constitute acceptable capability to withstand station blackout provided an analysis is performed which demonstrates that the plant has this capability from onset of the station blackout until the alternate AC source(s) and required shutdown equipment are started and lined up to operate. The time required for startup and alignment of the alternate AC power source(s) and this equipment shall be demonstrated by test. Alternate AC source(s) serving a multiple unit site where onsite emergency AC source are not shared between units must have, as a minimum, the capacity and capability for coping with a station blackout in any of the units. At sites where onsite emergency AC sources are shared between units, the alternate AC source(s) must have the capacity and capability as required to ensure that all units can be brought to and maintained in safe shutdown (non-DBA) as defined in § 50.2. If the alternate AC source(s) meets the above requirements and can be demonstrated by test to be available to power the shutdown buses within 10 minutes of the onset of station blackout, then no coping analysis is required.

(3) Regulatory Assessment:

(4) Implementation Schedule: (53 FR 23215, June 21, 1988)

The ABWR CTG will be automatically initiated upon the loss of power to the PIP buses. The CTG will achieve required speed and voltage within 2 minutes. The CTG will be manually connected to safe shutdown buses within 10 minutes. These equipment capabilities will be demonstrated 1) by the manufacturer’s component tests, 2) by the CTG initial startup tests and 3) periodically by the COL applicant as part of his operational reliability assurance program.

The ABWR design is a single unit plant arrangement design.

The CTG AAC source is available to power shutdown loads within 10 minutes as described above. Therefore, no coping analysis is required. In addition, the ABWR is designed with an 8-hour battery to accommodate station blackout without the need for AC power. Also, the three independent emergency diesel generator systems will accommodate one DG out of service, plus a single failure, with the remaining DG capable of bringing the plant to safe shutdown.



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Table 1C-2 ABWR Design Compliance with Regulatory Guide 1.155


Regulatory Guide 1.155—Station Blackout

Regulatory Position

3.3.5 If an AAC power source is selected specifically for satisfying the requirements for station blackout, the design should meet the following criteria:

1. The AAC power source should not normally be directly connected to the preferred or the blacked-out unit’s onsite emergency AC power system.

The ABWR AAC power source is not normally connected to the preferred or the onsite emergency AC power system. Two open circuit breakers—one Class 1E and the other non-Class 1E—separate the CTG from the safety-related emergency buses.

The AAC power source is also not normally connected to any of the preferred AC power sources or their associated non-safety-related buses. A non-Class 1E circuit breaker separates the CTG from the PIP buses.

2. There should be a minimum potential for common cause failure with the preferred or the blacked-out unit’s onsite emergency AC power sources. No single-point vulnerability should exist whereby a weather-related event or single active failure could disable any portion of the blacked-out unit’s onsite emergency AC power sources or the preferred power sources and simultaneously fail the AAC power source.

The ABWR design minimizes the potential for a) common cause failures between the preferred sources and the onsite emergency power sources; b) common cause failures between onsite emergency power sources themselves; c) common cause failures between onsite power sources and the AAC power source; and d) common cause failures between preferred sources and the AAC power source.

The design also precludes interactions between preferred, onsite emergency, and AAC power systems resulting from weather related events or single failures such that a single point vulnerability will not simultaneously fail both the AAC power source and the onsite emergency or offsite preferred power source(s). This is accomplished by having onsite emergency and the AAC power sources inside weather protected buildings and by maintaining adequate separation between the four power sources. None of the four standby power sources share emergency buses or loads, auxiliary services or instrumentation and controls prior to the recovery actions from the SBO event. These power sources are physically, electrically, mechanically and environmentally separated.

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3. The AAC power source should be available in a timely manner after the onset of station blackout and have provisions to be manually connected to one or all of the redundant safety buses as required. The time required for making this equipment available should not be more than 1 hour as demonstrated by test. If the AAC power source can be demonstrated by test to be available to power the shutdown buses within 10 minutes of the onset of station blackout, no coping analysis is required.

The ABWR AAC design power source will be automatically started and reach rated speed and voltage and be available to supply PIP loads within 2 minutes, and safety-related loads within 10 minutes for any loss of preferred offsite power sources (LOPP).

The design has provisions to assure the timely manual interconnection between the AAC (CTG) and any one or more of the safety-related shutdown buses.

The ABWR AAC design will be demonstrated by test to show that it can be connected to safety-related buses within 10 minutes. Therefore, no coping analysis is required.

4. The AAC power source should have sufficient capacity to operate the systems necessary for coping with a station blackout for the time required to bring and maintain the plant in safe shutdown.

The ABWR AAC power source is rated at 9 MWe, which is more than sufficient capacity to operate the necessary safe shutdown loads which are less than 5 MWe.

5. The AAC power system should be inspected, maintained, and tested periodically to demonstrate operability and reliability. The reliability of the AAC power system should meet or exceed 95% as determined in accordance with NSAC-108 (Reference 11) or equivalent methodology.

The ABWR design includes previsions to demonstrate the operability and reliability of the AAC power source. The CTG will be subject to surveillance inspection, testing and maintenance in accordance with the manufacturer’s requirements, the COL applicant’s maintenance program and with operational reliability assurance program requirements. The CTG will meet or exceed a reliability goal of 0.95 in accordance with NSAC-108 or equivalent methodology.

Table 1C-2 ABWR Design Compliance with Regulatory Guide 1.155 (Continued)


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3.3.6 If a system or component is added specifically to meet the recommendations on station blackout duration in Regulatory Position 3.1, system walk downs and initial tests of new or modified systems or critical components should be performed to verify that the modifications were performed properly. Failures of added components that may be vulnerable to internal or external hazards within the design basis (e.g., seismic events) should not affect the operation of systems required for the design basis accident.

The ABWR design includes the CTG as the AAC power source for SBO mitigation. A test program will be conducted by the manufacturer/equipment vendor to verify the major equipment performance objectives (e.g., start time, rated speed and voltage times, stable voltage outputs, etc.). These tests will be conducted prior to CTG installation at the plant site. Prior to plant operation, the AAC power source will be subject to pre-operational testing to demonstrate that the CTG will perform its intended function. Periodically, the AAC power source will be tested to assure that the reliability/availability goals are being met and maintained.

The ABWR design safety evaluations take into account potential plant disturbances that could affect AAC power source reliability. These disturbances could occur as a result of internal and external hazards (e.g., floods, fires and harsh environs, respectively). The adverse effects on AAC power source components due to operational hazards will not affect the operations of safety-related systems required for the design basis events. The effects caused by or upon the AAC power source due to operational events (internal and external hazards) are limited since the AAC power source components are physically, mechanically and essentially electrically isolated from the design basis engineered safety features and other power generation systems and components. Design bases accident events may result in the potential degradation of the AAC power source. However, the resulting effects of the AAC will not diminish the current safety system responses and the current event outcomes.



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3.3.7 The system or component added specifically to meet the recommendations on station blackout duration in Regulatory Position 3.1 should be inspected, maintained, and tested periodically to demonstrate equipment operability and reliability.

The ABWR design AAC power source will be capable of being tested, inspected and maintained on a periodic basis.

The CTG location in the Turbine Building provides easy access to the unit. The access and environmental conditions in the CTG area allow physical surveillance, easy maintenance, and testing.

The CTG will be periodically started, brought up to speed and voltage, and connected to the PIP buses.

The CTG will be subject to periodic test in order to verify the operability and reliability goals in the plant operational reliability assurance program (ORAP).

3.4 Procedures and Training To Cope with Station Blackout

Procedures* and training should include all operator actions necessary to cope with a station blackout for at least the duration determined according to Regulatory Position 3.1 and to restore normal long-term core cooling/decay heat removal once AC power is restored.

Appropriate plant procedures will be developed by the COL applicant for the ABWR design. These procedures will be integrated/coordinated with the plant EOPs, using the EOP methodology. Procedures will consider instructions for operator actions, responses, timing, and related matters during the SBO event.

* Procedures should be integrated with plant-specific technical guidelines and emergency procedures developed using the emergency operating procedure upgrade program established in response to Supplement 1 of NUREG-0737 (Reference 12). The task analysis portion of the emergency operating procedure upgrade program should include an analysis of instrumentation adequacy during a station blackout.



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3.5 Quality Assurance and Specification Guidance for Station Blackout Equipment that is Not Safety-Related

Appendices A and B provide guidance on quality assurance (QA) activities and specifications respectively for non-safety-related equipment used to meet the requirements of § 50.63 and not already covered by existing QA requirements in Appendix B or R of Part 50. Appropriate activities should be implemented from among those listed in these appendices depending on whether the non-safety equipment is being added (new) or is existing. This QA guidance is applicable to non-safety systems and equipment for meeting the requirements of § 50.63 of 10CFR50. The guidance on QA and specifications incorporates a lesser degree of stringency by eliminating requirements for involvement of parties outside the normal line organization. NRC inspections will focus on the implementation and effectiveness of the quality controls described in Appendices A and B. Additionally, the equipment installed to meet the station blackout rule must be implemented such that it does not degrade the existing safety-related systems. This is to be accomplished by making the non-safety-related equipment as independent as practicable from existing safety-related systems. The non-safety systems identified in Appendix B are acceptable to the NRC staff for responding to a station blackout.

The ABWR AAC power source design addresses the quality assurance and equipment specification guidance indicated in Appendices A and B of this guide.

The specific responses to Appendices A and B are presented in the following sections in this table.



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Appendix A – Quality Assurance

The QA guidance provided here is applicable to non-safety systems and equipment used to meet the requirements of § 50.63 and not already explicitly covered by existing QA requirements in 10CFR50 in Appendix B or R. Additionally, non-safety equipment installed to meet the station blackout rule must be implemented so that it does not degrade the existing safety-related systems. This is accomplished by making the non-safety equipment as independent as practicable from existing safety-related systems. The guidance provided in this section outlined an acceptable QA program for non-safety equipment used for meeting the station blackout rule and not already covered by existing QA requirements. Activities should be implemented from this section as appropriate depending on whether the equipment is being added (new) or is existing.

The ABWR AAC power source design is in compliance with the following QA guidelines in 10CFR50.63 as indicated below:

1. Design Control and Procurement Document Control

Measures should be established to ensure that all design-related guidances used in complying with §

50.63 are included in design and procurement documents, and that deviations therefrom are controlled.

The COL applicant’s QA program will comply with this requirement.

2. Instructions, Procedures, and Drawings

Inspections, tests, administrative controls, and training necessary for compliance with §

50.63 should be prescribed by documented instructions, procedures, and drawings and should be accomplished in accordance with these documents.


3. Control of Purchased Material, Equipment, and Services

Measures should be established to ensure that purchased material, equipment, and services conform to the procurement documents.




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4. Inspection

A program for independent inspection of activities required to comply with §

50.63 should be established and executed by (or for) the organization performing the activity to verify conformance with documented installation drawings and test procedures for accomplishing the activities.


5. Testing and Test Control

A test program should be established and implemented to ensure that testing is performed and verified by inspection and audit to demonstrate conformance with design and system readiness requirements. The tests should be performed in accordance with written test procedures; test results should be properly evaluated and acted on.


6. Inspection, Test, and Operating Status

Measures should be established to identify items that have satisfactorily passed required tests and inspections.


7. Nonconforming Items

Measures should be established to control items that do not conform to specified requirements to prevent inadvertent base or installation.


8. Corrective Action

Measures should be established to ensure that failures, malfunctions, deficiencies, deviations, defective components, and nonconformances are promptly identified, reported, and corrected.




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9. Records

Records should be prepared and maintained to furnish evidence that the criteria enumerated above are being met for activities required to comply with §

50.63.


10. Audits

Audits should be conducted and documented to verify compliance with design and procurement documents, instructions, procedures, drawings, and inspection and test activities developed to comply with § 50.63.




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Table 1C-2 ABWR Design Compliance with RG 1.155 (Continued)


Appendix B—Guidance Regarding Systems/Components

Alternate AC Sources ABWR AAC Power Source

Safety-RelatedEquipment(Compliance with IEEE-279)

Not required, but the existing Class 1E electrical systems must continue to meet all applicable safety-related criteria.

Existing onsite emergency power sources, buses and loads will continue to meet all applicable safety-related criteria.

Redundancy Not required. —

Diversity from Existing EDGs

See Regulatory Position 3.3.4 of this guide. The ABWR design will utilize a AAC diverse power source from that of the EDGs. A qualified combustion turbine generator will be used as the AAC.

Independencefrom Existing Safety-RelatedSystems

Required if connected to Class 1E buses. Separation to be provided by 2 circuit breakers in series (1 Class 1E at the Class 1E bus and 1 non-Class 1E).

Two breakers separate the onsite emergency power buses from the CTG. One breaker is Class 1E and the breaker closest to the CTG is non-Class 1E (see Figure 8.3-1).

SeismicQualification

Not required. —

EnvironmentalConsideration

If normal cooling is lost, needed for station blackout event only and not for design basis accident (DBA) conditions. Procedures should be in place to affect the actions necessary to maintain acceptable environmental conditions for the required equipment. See Regulatory Position 3.2.4.

The use of the ACC power source will assure that the plant equipment/environment cooling loss will be limited to 10 to 60 minutes (SBO duration). Normal plant cooling loads will be restored after shutdown loads are reestablished. Temperature rise conditions will be limited to minutes rather than hours

Capacity Specified in § 50.63 and Regulatory Position 3.3.4. The AAC power source is capable of powering more than the minimum required shutdown loads.

Quality Assurance Indicated in Regulatory Position 3.5. The ABWR design will be subjected to the quality assurance standards cited in Appendix A.

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Alternate AC Sources ABWR AAC Power Source

Technical Specification for Maintenance,Limiting Condition, FSAR, etc.

Should be consistent with the Interim Commission Policy Statement on Technical Specifications (Federal Register Notice 52 FR 3789) as applicable.

The AAC power source operational and test requirements will be defined by the Plant Maintenance Program and the ORAP. They will also be consistent with the Interim Commission Policy Statement on Tech Specs.

Instrumentationand Monitoring

Must meet system functional requirements. The AAC power source instrumentation, controls and monitoring will be of such number, type and quality to assure that the CTG reliability goals are met.

Single Failure Not required. —

Common Cause Failure (CCF)

Design should, to the extent practicable, minimize CCF between safety-related and non-safety-related equipment.

The AAC power source will be physically, mechanically and electrically independent of the offsite and onsite power systems.



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Water Source (Existing Condensate

Storage Tank or Alternate) SBO Recovery with AAC Power Source


Not required, but the existing Class 1E systems must continue to meet all applicable safety-related criteria.

The ABWR design Condensate Storage Tank will provide primary makeup water via the RCIC or HPCF. The suppression pool will serve as the secondary water source. The AAC powered RCWS and RSWS pumps will provide heat removal service to the plant systems including chillers and HVAC cooling subsystems.


Diversity Not required. —


Ensure that the existing safety functions are not compromised, including the capability to isolate components, subsystems, or piping, if necessary.

The loss of all AC power (SBO) will automatically cause reactor scram, MSIV closure, and initiation of the RCIC. The AAC power source will re-energize the lost shutdown loads (emergency makeup water, heat removal and HVAC services) due to the SBO condition within ten (10) to 60 minutes. The condensate storage tank will used during the first ten minutes and throughout the hot shutdown transition period. A significant amount of water is available from the CST (e.g. 2271 m3). After restoration of power via AAC other plant makeup and cooling water sources will be made available.


Not required. —


Need for station blackout event only and not for DBA conditions. See Regulatory Position 3.2.4. Procedures should be in place to effect the actions necessary to maintain acceptable environmental conditions for required equipment.

The AAC power source does not need plant service or cooling water for operation. It’s a self (air) cooled, self-lubricated and self-controlled machine.



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Capacity Capability to provide sufficient water for core cooling in the event of a station blackout for the specified duration to meet § 50.63 and this regulatory guide.

The Condensate Storage Tank (CST) is capable of providing at least 8 hours of makeup water without replenishment. With the use of the AAC power sources other water sources are readily available for makeup, heat removal, and plant equipment cooling.

Quality Assurance As indicated in Regulatory Position 3.5. The ABWR design’s immediate response to an SBO event does utilize a non-safety makeup water source (the CST). The AAC power source will allow the use of non-safety water sources.

Technical Specifications for Maintenance,Surveillance,LimitingConditions, FSAR, etc.


No additional non-safety-related water sources are required during the duration of the 10- to 60-minute SBO event. Use of other sources during cold shutdown activities is optional.


Must meet system functional requirements. The makeup water source instrumentation and controls, used during the SBO duration, are safety-related and divisionally separated.



Design should, to the extent practicable, minimize CCF between safety-related and non-safety-related systems.

The primary makeup water source (Condensate Storage Tank) and the secondary makeup water source (Suppression Pool), utilized during the 10 minute SBO duration, are physically, mechanically and environmentally separated from one another.



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Instrument Air

(Compressed Air System) SBO Recovery with AAC Power Source



Use of Plant Instrument Air/Compressed Air Systems during the 10 minute SBO duration is not required. Plant air systems availability is restored after 10 minutes by the AAC power source. Safety-related SRV nitrogen gas sources are available during the SBO event and are independent of non-safety air systems.





Air systems are not required to operate during the SBO duration. The CTG unit does not depend on an air starter system nor air supplied services. The CTG does have a self-contained intake and exhaust system. This is provided by the machine power sources itself.


Not required. —


Needed for station blackout event only and not for DBA conditions. See Regulatory Position 3.2.4. Procedures should be in place to effect the actions necessary to maintain acceptable environmental conditions for required equipment.

The CTG does not require special air or environmental control services before, during or after the SBO event.



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Capacity Sufficient compressed air to components, as necessary, to ensure that the core is cooled and appropriate containment integrity is maintained for the specified duration of station blackout to meet §

50.63 and Regulatory Guide 1.155.

Air service may be utilized later in the SBO recovery stage to reconfigure plant system to normal operation alignments.

Quality Assurance As indicated in Regulatory Position 3.3. Non-safety-related air systems are not utilized during the 10 minute SBO duration.



The CTG does not require air start services. The unit is started by a self-contained diesel engine starting system.


Must meet system functional requirements. Plant air system instrumentation, control and monitoring is not required during the 10 minute SBO duration.




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Water Delivery System (Alternative

to Auxiliary Feedwater System,

RCIC System, or Isolation

Condenser Makeup) SBO Recovery with AAC Power Source



The ABWR AAC power source design response during the 10 minute SBO duration does not require additional water makeup sources beyond the CST and/or the Suppression Pool.

Later in the SBO recovery sequence, the ABWR will utilize the normal plant water systems by powering selective divisions with the AAC power source (e.g. reactor service water and reactor cooling water systems).





The powering of the normal plant water sources by the AAC power source during SBO will not be inconsistent or contrary with their current DBA design basis.


Not required. —


Need for station blackout event only and not for DBA conditions. See Regulatory Position 3.2.4. Procedures should be in place to effect the actions necessary to maintain acceptable environmental conditions for required equipment.

The use of the normal plant cooling water systems will not require prior equipment environment controls or cooling. Their operation will be provided concurrently with the powering of water makeup sources.



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Water Delivery System (Alternative

to Auxiliary Feedwater System,

RCIC System, or Isolation

Condenser Makeup) SBO Recovery with AAC Power Source

Capacity Capability to provide sufficient water for core cooling in the event of a station blackout for the specified duration to meet § 50.63 and this regulatory guide.

The emergency water makeup sources include the condensate storage tank and the suppression pool inventory. The normal plant water makeup sources (component and service water, etc.) are in addition to other alternative core and containment makeup sources (e.g., feedwater, fire pumps, makeup water systems, etc.) all of these systems can supply makeup or cooling water.

Quality Assurance As indicated in Regulatory Position 3.5. The plant normal makeup water systems are subject to quality assurance evaluations (e.g. CST and the SP).



Emergency water makeup systems are subject to Technical Specifications requirements.


Must meet system functional requirements. Instrumentation and controls for normal plant makeup water systems are qualified for their functional services.




The use of additional plant water makeup systems (post SBO) will not degrade the operation or reliability of the necessary makeup systems (RCIC, HPCF, etc.). The CTG has sufficient capacity to power necessary shutdown loads and selective other safety and non-safety loads needed for water makeup.



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Table 1C-3 ABWR Design Compliance with NUMARC 87-00 Guidelines


7.0 Coping Evaluations

7.1.1 Coping Methods

For purposes of this assessment, coping methods are separated into two different approaches. The first is referred to as the “AC-Independent” approach. In this approach, plants rely on available process steam, DC power, and compressed air to operate equipment necessary to achieve safe shutdown conditions (i.e., Hot Standby or Hot Shutdown, as appropriate) until offsite or emergency AC power is restored. A second approach is called the “Alternate AC” approach. This method is named for its use of equipment that is capable of being electrically isolated from the preferred offsite and emergency onsite AC power sources. Station blackout coping using the Alternate AC power approach would entail a short period of time in an AC-Independent state (up to one hour) while the operators initiate power from the backup source. Once power is available, the plant would transition to the Alternate AC state and provide decay heat removal until offsite or emergency AC power becomes available. The AC power sources used in the Alternate AC power approach would be subject to the Appendix B criteria including electrical isolation requirements in order to assure their availability in the event of a station blackout.

Appendix A provides a definition of Alternate AC power sources. Appendix B provides detailed acceptance criteria for an Alternate AC power source.

The ABWR design utilizes the “Alternate AC (AAC)” approach as defined in Appendix A. The AAC power source will be available to be connected to the core inventory makeup and decay heat removal loads within ten (10) minutes. The AAC power source is capable of being electrically isolated from the preferred offsite and emergency onsite AC power sources and complies with the Appendix B criteria including electrical isolation requirements.

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7.1.2 Coping Duration

AC-Independent plants must meet the requirements of this methodology for at least four hours (or at least two hours for plants in both emergency AC group A and offsite power group P1). Plants using an Alternate AC power source must assess their ability to cope for one hour. However, if an Alternate AC power source can be shown by test to be available within 10 minutes of the onset of station blackout, then no coping assessment is required. Available within 10 minutes means that circuit breakers necessary to bring power to safe shutdown buses are capable of being actuated in the control room within that period.

ABWR design will demonstrate by test that the AAC CTG is capable of being available within ten (10) minutes of the onset of a SBO event and therefore no formal coping evaluation is necessary or required. All actions during the 10 minute period are safety-related and automatic. The ABWR design provides the operator in the main control room with the means to reconfigure the electrical distribution system including circuit breakers, and to connect the AAC power source to the necessary shutdown buses and loads within the ten (10) minute interval.

Table 1C-3 ABWR Design Compliance with NUMARC 87-00 Guidelines (Continued)


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Appendix A — Definitions

This appendix defines the terminology used throughout the guide.

ALTERNATE AC POWER SOURCE. An alternating current (AC) power source that is available to and located at or nearby a nuclear power plant and meets the following requirements:

(i) Is connectable to but not normally connected to the preferred or onsite emergency AC power systems

(ii) Has minimal potential for common cause failure with offsite power or the onsite AC power sources

(iii) Is available in a timely manner after the onset of station blackout

(iv) Has sufficient capacity and reliability for operation of all systems necessary for coping with a station blackout and for the time required to bring and maintain the plant in safe shutdown (Hot Shutdown or Hot Standby, as appropriate)

(v) Is inspected, maintained, and tested periodically to demonstrate operability and reliability as set forth in Appendix B

The ABWR AAC power source design will meet the following requirements:

(i) The design is connectable to (but not normally connected to) the preferred or onsite emergency AC power sources. Two normally open breakers separate the AAC CTG from the safety-related onsite emergency power buses. A single normally open breaker separates the AAC CTG from the non-safety-related PIP buses (preferred power) (see Figure 8.3-1).

(ii) The ABWR design has a minimal potential for common cause failure between preferred power or onsite AC power sources. The ABWR AAC power source is a diverse power supply to the normal onsite emergency DGs. The AAC power supply is totally independent of the preferred and onsite power sources. The AAC power source automatically starts and is available for loading in two minutes. The AAC power supply is connectable to a Class 1E bus through the actuation of two (2) manual operated circuit breakers. The AAC power source is normally electrically, physically, mechanically, and environmentally isolated from the preferred and onsite power sources. The AAC power source is normally used during LOPP and SBO events. However, the CTG can be used for a number of operational services (e.g. maintenance backup, etc.).

(iii) The ABWR AAC power source is available in a timely manner after the onset of a SBO event. The AAC power source automatically starts on LOPP, attains required speed and voltage within two (2) minutes, and is capable of being connected to shutdown loads within ten (10) minutes.

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(iv) The ABWR AAC power source is rated a minimum of 9 MWe. The shutdown loads are less than 5 MWe. The CTG reliability is 0.95. The ABWR is expected to be in hot shutdown condition in twenty four (24) hours, and in cold shutdown condition in ninety-six (96) hours. The CTG, is designed to run indefinitely under SBO conditions at rated load. A seven-day fuel supply is available on the site for the CTG.

(v) The ABWR AAC power source will be capable of being inspected, maintained and tested periodically to demonstrate its operability and reliability to guidelines set forth in Appendix B.

REQUIRED COPING DURATION. The time between the onset of station blackout and the restoration of offsite AC power to safe shutdown buses.

The ABWR AAC power source design does not require a formal SBO coping analysis. The AAC power source will be available to supply shutdown loads within ten (10) minutes. The current design requirements associated with DBA events assure that the plant will be able to cope with a ten (10) minute SBO event.

SAFE SHUTDOWN. For the purpose of this procedure safe shutdown is the plant conditions defined in plant technical specifications as Hot Standby or Hot Shutdown, as appropriate.

The ABWR design will assure safe shutdown plant conditions as defined by the Plant Technical Specifications and the definition in 10CFR50.63.

STATION BLACKOUT. Means the complete loss of alternating current (AC) electric power to the essential and nonessential switchgear buses in a nuclear power plant (i.e., loss of offsite electric power system concurrent with turbine trip and unavailability of onsite emergency AC power system). Station Blackout does not include the loss of available AC power to buses fed by station batteries through inverters or by Alternate AC power sources as defined in this appendix, nor does it assume a concurrent single failure or a design basis accident. At a multi-unit site, station blackout is assumed to occur in only one unit unless the emergency AC power sources are totally shared between the units.

The ABWR design accommodates the SBO definition and the other definitions defined in 10CFR50.63. The ABWR design utilizes the current available station batteries throughout the event. The station batteries will be recharged as necessary by the AAC power source.



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Appendix B—Alternate AC Power Criteria

This appendix describes the criteria that must be met by a power supply in order to be classified as an Alternate AC power source. The criteria focus on ensuring that station blackout equipment is not unduly susceptible to dependent failure by establishing independence of the AAC system from the emergency and non-Class 1E AC power systems.

AAC Power Source Criteria

B.1 The AAC system and its components need not be designed to meet Class 1E or safety system requirements. If a Class 1E EDG is used as an Alternate AC power source, this existing Class 1E EDG must continue to meet all applicable safety-related criteria.

The ABWR AAC power source is a non-safety-related CTG.

B.2 Unless otherwise provided in this criteria, the AAC system need not be protected against the effects of:

The ABWR AAC power source is housed in a Uniform Building Code Building (Turbine Building). The AAC power source is physically, mechanically, electrically and environmentally separated from the preferred and onsite power sources. The AAC power source is protected from normal plant and site environmental perturbations (e.g., wind, temperature, etc.).

(a) Failure or misoperation of mechanical equipment, including (i) fire, (ii) pipe whip, (iii) jet impingement, (iv) water spray, (v) flooding from a pipe break, (vi) radiation, pressurization, elevated temperature or humidity caused by high or medium energy pipe break, and (vii) missiles resulting from the failure of rotating equipment or high energy systems

(b) Seismic events

B.3 Components and subsystems shall be protected against the effects of likely weather-related events that may initiate the loss of offsite power event. Protection may be provided by enclosing AAC components within structures that conform with the Uniform Building Code, and burying exposed electrical cable run between buildings (i.e., connections between the AAC power source and the shutdown buses).

The ABWR AAC power source is protected against the effects of weather-related events that may initiate the loss of offsite power events. The AAC power source is located above the maximum flood level in the Turbine Building. The power and control cables from the CTG to the shutdown buses are routed separately from the offsite preferred power and control cables to the shutdown buses in the Reactor Building. The Turbine Building design basis capabilities will provide adequate protection for the enclosed equipment in compliance with their equipment design basis requirements.

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B.4 Physical separation of AAC components from safety-related components or equipment shall conform with the separation criteria applicable for the unit’s licensing basis.

The ABWR AAC power source design maintains physical separation between safety-related components or equipment and the CTG by adhering to applicable separation criteria used in the plant licensing basis.

Connectability to AC Power Systems

B.5 Failure of AAC components shall not adversely affect Class 1E AC power systems.

The ABWR AAC power source design and its associated components failures will not adversely affect Class 1E AC power systems. Class 1E AC power system failures will not affect AAC power source operability.

B.6 Electrical isolation of AAC power shall be provided through an appropriate isolation device. If the AAC source is connected to Class 1E buses, isolation shall be provided by two circuit breakers in series (one Class 1E breaker at the Class 1E bus and one non-Class 1E breaker to protect the source).

The ABWR AAC power source is electrically isolated from the Class 1E power sources by two (2) circuit breakers in series (one Class 1E at the Class 1E buses and one non-Class 1E breaker at the CTG bus). Power to the breakers will be from appropriate DC sources.

B.7 The AAC power source shall not normally be directly connected to the preferred or onsite emergency AC power system for the unit affected by the blackout. In addition, the AAC system shall not be capable of automatic loading of shutdown equipment from the blacked-out unit unless licensed with such capability.

The ABWR AAC power source will not normally be connected to the preferred or onsite emergency AC power system. However, the COL applicant may use the CTG for other services (e.g. maintenance backup, etc.). The AAC power system will not automatically connect to or load any shutdown equipment on safety-related emergency buses. The AAC power source will automatically start upon occurrence of a LOPP event. It is connected automatically to the non-safety-related Plant Investment Protection (PIP) buses. It is capable of being manually connected to safety-related buses. It is also capable of being manually connected to non-safety power generation loads, condensate pumps, etc.).

Minimum Potential for Common Cause Failure

B.8 There shall be minimal potential for common cause failure of the AAC power source(s). The following system features provide assurance that the minimal potential for common cause failure has been adequately addressed.

The ABWR AAC power source design contains a number of design and operational features which provide assurance of minimal potential for common cause failure.



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(a) The AAC power system shall be equipped with a DC power source that is electrically independent from the blacked-out unit’s preferred and Class 1E power system.

The AAC power system is equipped with sufficient plant or self-contained non-Class 1E DC power supplies (separate from the Class 1E DC power supplies) to facilitate successful operation.

During normal operation, the plant electrical distribution systems will provide charging power to the plant battery systems.

(b) The AAC power system shall be equipped with an air start system, as applicable, that is independent of the preferred and the blacked-out unit’s preferred and Class 1E power supply.

The AAC power system is equipped with a self-contained, independent diesel engine hydraulic starting system. This starter is designed for SBO conditions. The entire starter assembly is mounted on the same skid with the CTG.

(c) The AAC power system shall be provided with a fuel oil supply, as applicable, that is separate from the fuel oil supply for the onsite emergency AC power system. A separate day tank supplied from a common storage tank is acceptable provided the fuel oil is sampled and analyzed consistent with applicable standards prior to transfer to the day tank.

The AAC power supply is equipped with a fuel system separate from that of the DGs. An external fuel supply transfer system will also be provided. A seven (7) day supply of oil for use by the CTG to achieve safe shutdown is available on site. The CTG oil storage and transfer system is physically and mechanically independent of the DG oil storage and transfer system.

(d) If the AAC power source is an identical machine to the emergency onsite AC power source, active failures of the emergency AC power source shall be evaluated for applicability and corrective action taken to reduce subsequent failures.

The ABWR AAC power source is an independent and diverse power supply from the onsite emergency DG power sources. The AAC power source is a combustion turbine generator.

(e) No single point vulnerability shall exist whereby a likely weather-related event or single active failure could disable any portion of the onsite emergency AC power sources or the preferred power sources, and simultaneously fail the AAC power source(s).

The ABWR of the AAC power source design precludes single point vulnerabilities, weather-related events effects, or single active failures that could disable any portion of the onsite emergency AC power sources or the preferred power sources and simultaneously fail the AAC power source.

The AAC power source is physically, mechanically, electrically and environmentally separated from the other plant power systems (e.g. circuit breaker separation, separate oil supplies, separate auto start circuits, etc.).



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(f) The AAC power system shall be capable of operating during and after a station blackout without any support systems powered from the preferred power supply, or the blacked-out unit’s Class 1E power source affected by the event.

The ABWR AAC power source design does not require preferred or onsite AC power sources to support the operation of the CTG unit. The CTG and its auxiliary support systems are maintained in their standby status by normal plant power sources.

Upon reaching design speed and voltage, the CTG operation is supported by a self-powered internal control package. This package assures continued operation without external power or auxiliary service needs.

(g) The portions of the AAC power system subjected to maintenance activities shall be tested prior to returning the AAC power system to service.

The ABWR AAC power source is capable of being tested and will be periodically tested:

(i) To demonstrate its reliability and its availability

(ii) To demonstrate that it can be connected to shutdown buses within ten (10) minutes from the MCR

(iii) To demonstrate the operability after maintenance has been performed on the CTG

Availability After Onset of Station Blackout

B.9 The AAC power system shall be sized to carry the required shutdown loads for the required coping duration determined in Section 3.2.5, and be capable of maintaining voltage and frequency within limits consistent with established industry standards that will not degrade the performance of any shutdown system or component. At a multi-unit site, except for 1/2 shared or 2/3 emergency AC power configurations, an adjacent unit’s Class 1E power source may be used as an AAC power source for the blacked-out unit if it is capable of powering the required loads at both units.

The ABWR AAC power source is designed to provide reliable power to shutdown loads during and after the SBO duration. The CTG will maintain supply voltage and frequency within the limits currently required during normal operation, and during loading transients, etc.



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Capacity and Reliability

B.10 Unless otherwise governed by technical specifications, the AAC power source shall be started and brought to operating conditions that are consistent with its function as an AAC source at intervals not longer than three months, following manufacturer’s recommendations or in accordance with plant-developed procedures. Once every refueling outage, a timed start (within the time period specified under blackout conditions) and rated load capacity test shall be performed.

The ABWR AAC power source will be started and brought to operating conditions consistent with manufacturer’s recommendations, the plant ORAP, or in accordance with specific plant developed procedures. This is a COL applicant interface item.

The AAC power source is capable of being started and connected to the preferred power source for load capacity testing.

The COL applicant will provide testing procedures based on plant specific ORAP objectives.

B.11 Unless otherwise governed by technical specifications, surveillance and maintenance procedures for the AAC system shall be implemented considering manufacturer’s recommendations or in accordance with plant-developed procedures.

Plant specific surveillance and maintenance procedures based on the appropriate manufacturer’s/vendor’s recommendations, operational reliability assurance programs, plant maintenance effectiveness programs and plant operational requirements will be provided by the COL applicant.

B.12 Unless otherwise governed by technical specifications, the AAC system shall be demonstrated by initial test to be capable of powering required shutdown equipment within one hour of a station blackout event.

The ABWR AAC power source design will be tested to demonstrate that the CTG is capable of powering shutdown equipment within 10 minutes of the SBO event.



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B.13 The Non-Class 1E AAC system should attempt to meet the target reliability and availability goals specified below, depending on normal system state. In this content, reliability and availability goals apply to the overall AAC system rather than individual machines, where a system may comprise more than one AAC power source.

The ABWR AAC power source satisfies the following reliability and availability goal:

System reliability will be maintained at or above 0.95 per demand as determined in accordance with NSAC-108 methodology or its equivalent.

Periodic testing and maintenance, to assure this reliability, will be performed.

(a) Systems Not Normally Operated (Standby Systems)

System reliability should be maintained at or below 0.95 per demand, as determined in accordance with NSAC-108 methodology (or equivalent).

(b) Systems Normally operated (Online Systems)

Availability: AAC systems normally online should attempt to be available to its associated unit at least 95% of the time the reactor is operating.

Reliability: No reliability targets or standards are established for online systems.



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